WorldWideScience

Sample records for digital reactor control

  1. Digital control of research reactors

    International Nuclear Information System (INIS)

    Crump, J.C. III.; Richards, W.J.; Heidel, C.C.

    1991-01-01

    Research reactors provide an important service for the nuclear industry. Developments and innovations used for research reactors can be later applied to larger power reactors. Their relatively inexpensive cost allows research reactors to be an excellent testing ground for the reactors of tomorrow. One area of current interest is digital control of research reactor systems. Digital control systems offer the benefits of implementation and superior system response over their analog counterparts. At McClellan Air Force Base in Sacramento, California, the Stationary Neutron Radiography System (SNRS) uses a 1,000-kW TRIGA reactor for neutron radiography and other nuclear research missions. The neutron radiography beams generated by the reactor are used to detect corrosion in aircraft structures. While the use of the reactor to inspect intact F-111 wings is in itself noteworthy, there is another area in which the facility has applied new technology: the instrumentation and control system (ICS). The ICS developed by General Atomics (GA) contains several new and significant items: (a) the ability to servocontrol on three rods, (b) the ability to produce a square wave, and (c) the use of a software configurator to tune parameters affected by the actual reactor core dynamics. These items will probably be present in most, if not all, future research reactors. They were developed with increased control and overall usefulness of the reactor in mind

  2. Digital control system of advanced reactor

    International Nuclear Information System (INIS)

    Peng Huaqing; Zhang Rui; Liu Lixin

    2001-01-01

    This article produced the Digital Control System For Advanced Reactor made by NPIC. This system uses Siemens SIMATIC PCS 7 process control system and includes five control system: reactor power control system, pressurizer level control system, pressurizer pressure control system, steam generator water level control system and dump control system. This system uses three automatic station to realize the function of five control system. Because the safety requisition of reactor is very strict, the system is redundant. The system configuration uses CFC and SCL. the human-machine interface is configured by Wincc. Finally the system passed the test of simulation by using RETRAN 02 to simulate the control object. The research solved the key technology of digital control system of reactor and will be very helpful for the nationalization of digital reactor control system

  3. Using the digital reactor control systems at NPP

    International Nuclear Information System (INIS)

    Schirl, G.; Hertel, J.

    2006-01-01

    A conception of application of the digital reactor control systems (RCS) at NPP is presented. The digital RCS architecture and safety ensuring are considered. The strategy and algorithm of the operating NPP equipping with the new digital RCS are given too [ru

  4. A digital controller for the Omega West Reactor

    International Nuclear Information System (INIS)

    Minor, M.M.; Kaufman, M.D.; Smith, T.W.

    1992-05-01

    A new nuclear reactor control system for the Omega West Reactor (OWR) has been designed to replace the aging and hard to maintain controller presently installed. The controller uses single board computers, digital and analog input and output modules, and stepping motor indexers installed on a standard bus (VME bus). The eight poison control rod drive motors are replaced with stepping motors. The control algorithm for the OWR was not changed in order to expedite approval for installation. This report presents the results of the development of the new control system. Included in the report are copies of some of the software that drives the new controller

  5. Centralized digital computer control of a research nuclear reactor

    International Nuclear Information System (INIS)

    Crawford, K.C.

    1987-01-01

    A hardware and software design for the centralized control of a research nuclear reactor by a digital computer are presented, as well as an investigation of automatic-feedback control. Current reactor-control philosophies including redundancy, inherent safety in failure, and conservative-yet-operational scram initiation were used as the bases of the design. The control philosophies were applied to the power-monitoring system, the fuel-temperature monitoring system, the area-radiation monitoring system, and the overall system interaction. Unlike the single-function analog computers currently used to control research and commercial reactors, this system will be driven by a multifunction digital computer. Specifically, the system will perform control-rod movements to conform with operator requests, automatically log the required physical parameters during reactor operation, perform the required system tests, and monitor facility safety and security. Reactor power control is based on signals received from ion chambers located near the reactor core. Absorber-rod movements are made to control the rate of power increase or decrease during power changes and to control the power level during steady-state operation. Additionally, the system incorporates a rudimentary level of artificial intelligence

  6. Development of Reactor Protection System (RPS) in Reactor Digital Instrumentation and Control System (ReDICS)

    International Nuclear Information System (INIS)

    Mohd Khairulezwan Abdul Manan; Mohd Sabri Minhat; Ridzuan Abdul Mutalib

    2013-01-01

    RTP Research Reactor are in the process upgraded from analogue control console system to a digital control console system . Upgrade process requires a statistical study to improve safety during reactor operation. RPS was developed to meet the needs of operational safety and at the same time comply with the guidelines set by the IAEA. RPS is in analog and hardware with industry standard interfaced with digital DAC (Data Acquisition and Control) and OWS (Operator Work Station). (author)

  7. MAPLE-X10 reactor digital control system

    International Nuclear Information System (INIS)

    Deverno, M.T.; Hinds, H.W.

    1991-10-01

    The MAPLE-X10 reactor, currently under construction at the Chalk River Laboratories of Atomic Energy of Canada Limited, is a 10 MW t , pool-type, light-water reactor. It will be used for radioisotope production and silicon neutron transmutation doping. The reactor is controlled by a Digital Control System (DCS) and protected against abnormal process events by two independent safety systems. The DCS is an integrated control system used to regulate the reactor power and process systems. The safety philosophy for the control system is to minimize unsafe events arising from system failures and operational errors. this is achieved through redundancy, fail-safe design, automatic fault detection, and the selection of highly reliable components. The DCS provides both computer-controlled reactor regulation from the shutdown state to full power and automated reactor shutdown if safe limits are exceeded or critical sensors malfunction. The use of commercially available hardware with enhanced quality assurance makes the system cost effective while providing a high degree of reliability

  8. The digital reactor protection system for the instrumentation and control of reactor TRIGA PUSPATI (RTP)

    International Nuclear Information System (INIS)

    Nurfarhana Ayuni Joha; Izhar Abu Hussin; Mohd Idris Taib; Zareen Khan Abdul Jalil Khan

    2010-01-01

    Reactor Protection System (RPS) is important for Reactor Instrumentation and Control System. The RPS comprises all redundant electrical devices and circuitry involved in the generation of those initiating signals associated to the trip protective function. The instrumentation system for the RPS provides automatic protection signals against unsafe and improper reactor operation. The physical separation is provided for all of the redundant instrumentation systems to preserve redundancy. The safety protection systems using circuits composed of analog instruments and relays with relay contacts is difficult to realize from various reasons. Therefore, an application of digital technology can be said a logical conclusion also in the light of its functional superiority. (author)

  9. Design for the human-machine interface of a digitalized reactor control-room

    International Nuclear Information System (INIS)

    Qu Ronghong; Zhang Liangju; Li Duo; Yu Hui

    2005-01-01

    Digitalized technology is implemented in the instrumentation and control system of an in-construction research reactor, which advances information display in both contents and styles in a nuclear reactor control-room, and greatly improves human-machine interface. In the design for a digitalized nuclear reactor control-room there are a series of new problems and technologies should be considered seriously. This paper mainly introduces the design for the digitalized control-room of the research nuclear reactor and covered topics include design principle of human-machine interface, organization and classification of interface graphics, technologies and principles based on human factors engineering and implemented in the graphics design. (authors)

  10. Digital, remote control system for a 2-MW research reactor

    International Nuclear Information System (INIS)

    Battle, R.E.; Corbett, G.K.

    1988-01-01

    A fault-tolerant programmable logic controller (PLC) and operator workstations have been programmed to replace the hard-wired relay control system in the 2-MW Bulk Shielding Reactor (BSR) at Oak Ridge National Laboratory. In addition to the PLC and remote and local operator workstations, auxiliary systems for remote operation include a video system, an intercom system, and a fiber optic communication system. The remote control station, located at the High Flux Isotope Reactor 2.5 km from the BSR, has the capability of rector startup and power control. The system was designed with reliability and fail-safe features as important considerations. 4 refs., 3 figs

  11. Upgrading the Reactor Power Control Concept with a Modern Digital Control System

    International Nuclear Information System (INIS)

    Laengle, M.; Schildheuer, R.

    2011-01-01

    Within the framework of a retrofit project, a reactor power instrumentation and control system (REALL) - consisting of a limiting system and the respective reactor control systems - was retrofitted and modernized in a 1450-MW-nuclear-power-plant in Baden-Wuerttemberg. The REALL process control functions were implemented within a modern and completely digitized control system that has been designed for use in safety I and C applications. Along with the installation of the digital control system, the associated hardware was adapted to today's state of the art. At the same time, the given potential for improvement, as revealed during the plant's operation so far, was taken into account in the programming. In order to provide for transparent and quality-assured project management, the implementation was based on a stage plan consisting of several steps, along with specific milestones. Final commissioning of the modern digital control system took place during the 2008 plant overhaul. Despite the complex commissioning procedure, it was possible to avoid a major prolongation of the plant's downtime and to keep within a rough 4-week timeframe that had originally been defined for the plant overhaul to adequate structuring of the project, goal-oriented implementation of preparatory infrastructural measures and adequate scheduling of the coordinated activities of the installation and commissioning teams entrusted with the commissioning of the digital control system during the overhaul activities. (author)

  12. An automatic regulating control system for a graphite moderated reactor using digital techniques

    International Nuclear Information System (INIS)

    Carvalho Goncalves Junior, J. de.

    1989-01-01

    The work propose an automatic regulating control system for a graphite moderated reactor using digital techniques. The system uses a microcomputer to monitor the power and the period, to run the control algorithm, and to generate electronic signals to excite the motor, which moves vertically the control rod banks. A nuclear reactor simulator was developed to test the control system. The simulator consists of a software based on the point kinetic equations and implanted in an analogical computer. The results show that this control system has a good performance and versatility. In addition, the simulator is capable of reproducing with accuracy the behavior of a nuclear reactor. (author)

  13. Measures to enhance the reliability of digital reactor instrumentation and control system

    International Nuclear Information System (INIS)

    Li Duo; Zhang Liangju

    2005-01-01

    The instrumentation and control (I and C) system of an in-constriction research reactor is fully based on digital technology, which greatly improves system performance in many aspects such as MMI, data processing accurately, centralized control and monitoring ability, and maintainability. The reliability should be an essential issue for the successfulness of a digital I and C. This paper describes measures taken in the design of the reactor I and C to enhance the overall reliability, including distributed data processing; and control functions, redundancy, autonomous control, and diversified MMI, which would greatly contribute to the system reliability. (authors)

  14. Experimental direct digital control of the power plant A1 reactor based on a modern control theory approach

    International Nuclear Information System (INIS)

    Karpeta, C.

    1979-01-01

    The objective of the project was to accumulate technical experience with application of modern control theory in nuclear power by carrying out a case study of an experimental direct digital control at the A1 reactor about its nominal steady state. The research has proved that slightly modified method of solution of the linear stochastic regulator problem can be successfully applied in design of digital control system of nuclear power reactors

  15. Closed-loop digital control of nuclear reactors characterized by spatial dynamics

    International Nuclear Information System (INIS)

    Bernard, J.A.; Henry, A.F.; Lanning, D.D.; Meyer, J.E.

    1991-03-01

    This report describes the theoretical development and the evaluation via both simulation and, to a lesser degree, experiment of a digital method for the closed-loop control of power and temperature in reactors characterized by spatial dynamics. The major conclusions of the research are that (1) the sophistication of advanced reactor physics and thermal-hydraulic nodal methods is now such that accurate, real-time models of spatially-dependent, heterogeneous reactor cores can be run on present-generation minicomputers; (2) operation of both present-day commercial reactors as well as the multi-modular reactors now being considered for construction in the United States could be significantly improved by incorporating model-generated information on in-core conditions in a digital controller; and (3) digital controllers for spatially-dependent reactors should have a hierarchical or multi-tiered structure consisting of supervisory algorithms that preclude challenges to the safety system, global control laws designed to provide an optimal response to temperature and power perturbations, and local control laws that maintain parameters such as the margin to departure from nucleate boiling within specification. The technology described is appropriate to present-day pressurized water reactors and to the proposed multi-modular designs. The end-product of this research was a (near) real-time analytic plant-estimation code that was given the acronym POPSICLE for POwer Plant SImulator and ControlLEr. POPSICLE's core neutronics model is based on a quasi-static transient solution of the analytic nodal diffusion equations. 126 refs., 159 figs., 17 tabs

  16. Closed-loop digital control of nuclear reactors characterized by spatial dynamics

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, J.A. (Massachusetts Inst. of Tech., Cambridge, MA (USA). Nuclear Reactor Lab.); Henry, A.F.; Lanning, D.D.; Meyer, J.E. (Massachusetts Inst. of Tech., Cambridge, MA (USA). Dept. of Nuclear Engineering)

    1991-03-01

    This report describes the theoretical development and the evaluation via both simulation and, to a lesser degree, experiment of a digital method for the closed-loop control of power and temperature in reactors characterized by spatial dynamics. The major conclusions of the research are that (1) the sophistication of advanced reactor physics and thermal-hydraulic nodal methods is now such that accurate, real-time models of spatially-dependent, heterogeneous reactor cores can be run on present-generation minicomputers; (2) operation of both present-day commercial reactors as well as the multi-modular reactors now being considered for construction in the United States could be significantly improved by incorporating model-generated information on in-core conditions in a digital controller; and (3) digital controllers for spatially-dependent reactors should have a hierarchical or multi-tiered structure consisting of supervisory algorithms that preclude challenges to the safety system, global control laws designed to provide an optimal response to temperature and power perturbations, and local control laws that maintain parameters such as the margin to departure from nucleate boiling within specification. The technology described is appropriate to present-day pressurized water reactors and to the proposed multi-modular designs. The end-product of this research was a (near) real-time analytic plant-estimation code that was given the acronym POPSICLE for POwer Plant SImulator and ControlLEr. POPSICLE's core neutronics model is based on a quasi-static transient solution of the analytic nodal diffusion equations. 126 refs., 159 figs., 17 tabs.

  17. New digital control system for the operation of the Colombian research reactor IAN-R1

    International Nuclear Information System (INIS)

    Celis del A, L.; Rivero, T.; Bucio, F.; Ramirez, R.; Segovia, A.; Palacios, J.

    2015-09-01

    En 2011, Mexico won the Colombian international tender for the renewal of instrumentation and control of the IAN-R1 Reactor, to Argentina and the United States. This paper presents the design criteria and the development made for the new digital control system installed in the Colombian nuclear reactor IAN-R1, which is based on a redundant and diverse architecture, which provides increased availability, reliability and safety in the reactor operation. This control system and associated instrumentation met all national export requirements, with the safety requirements established by the IAEA as well as the requirements demanded by the Colombian Regulatory Body in nuclear matter. On August 20, 2012, the Colombian IAN-R1 reactor reached its first criticality controlled with the new system developed at Instituto Nacional de Investigaciones Nucleares (ININ). On September 14, 2012, the new control system of the Colombian IAN-R1 reactor was officially handed over to the Colombian authorities, this being the first time that Mexico exported nuclear technology through the ININ. Currently the reactor is operating successfully with the new control system, and has an operating license for 5 years. (Author)

  18. Undergraduate reactor control experiment

    International Nuclear Information System (INIS)

    Edwards, R.M.; Power, M.A.; Bryan, M.

    1992-01-01

    A sequence of reactor and related experiments has been a central element of a senior-level laboratory course at Pennsylvania State University (Penn State) for more than 20 yr. A new experiment has been developed where the students program and operate a computer controller that manipulates the speed of a secondary control rod to regulate TRIGA reactor power. Elementary feedback control theory is introduced to explain the experiment, which emphasizes the nonlinear aspect of reactor control where power level changes are equivalent to a change in control loop gain. Digital control of nuclear reactors has become more visible at Penn State with the replacement of the original analog-based TRIGA reactor control console with a modern computer-based digital control console. Several TRIGA reactor dynamics experiments, which comprise half of the three-credit laboratory course, lead to the control experiment finale: (a) digital simulation, (b) control rod calibration, (c) reactor pulsing, (d) reactivity oscillator, and (e) reactor noise

  19. New digital control and power protection system of VR 1 training reactor

    International Nuclear Information System (INIS)

    Kropik, M.; Matejka, K.; Juoeickova, M.

    2005-01-01

    The contribution describes the new VR-1 training reactor control and power protection system at the Czech Technical University in Prague. The control system provides safety and control functions, calculates average values of the important variables and sends data and system status to the human-machine interface. The upgraded control system is based on a high quality industrial PC. The operating system of the PC is the Microsoft Windows XP with the real time support RTX of the VentureCom Company. The software was developed according to requirements in MS Visual C. The independent power protection system is a component of the reactor safety (protection) system with high quality and reliability requirements. The digital system is redundant; each channel evaluates the reactor power and the velocity of power changes and provides safety functions. The digital part of the channel is multiprocessor-based. The software was developed with respect to nuclear standards. The software design was coded in the C language regarding the NRC restrictions. Configuration management, verification and validation accompanied the software development. Both systems were thoroughly tested. Firstly, the non active tests were carried out. During these tests, the active core of the reactor was subcritical; the input signals were generated from HPIB and VXI controlled instruments to simulate different operational and safety events. The software for instruments control and tests evaluation utilized Agilent VEE development system. After the successful non active checking, the active tests followed. (author)

  20. Micro controller based design of digital transmitters for temperature measurements in reactors

    International Nuclear Information System (INIS)

    Nassar, M.A.M.

    2011-01-01

    Temperature transmitter is one of the most important transmitters in the nuclear reactor it is used for RTD (resistance temperature detector) signal conditioning. It has built-in current excitation, instrumentation amplifier, linearization and current output circuitry which amplifies the RTD signal and gives linearization to it. It is a part of a system to get temperature and monitoring it. This system is very cost and complicated. In this work a digital system is implemented by using micro controller techniques that replaces the existing system, one chip (PIC16f877) is used to build a digital system, which is more accurate and give more performance and low costs . RTD is the sensing element of temperature, its resistance increases with temperature. There are many types of transmitters in the reactor such as temperature, pressure, level and flow but temperature one is chosen because of temperature is one of the most important parameters in process control.

  1. Performance analysis of the closed digital control circuit of reactor A-1

    International Nuclear Information System (INIS)

    Karpeta, C.; Volf, K.; Stirsky, P.; Roubal, S.; Muellerova, H.

    A computer-aided analysis is presented of the optimum digital control of the A-1 nuclear power plant reactor. The effect of index weighting matrices on the quality of control processes was studied for a deterministic case using the Separation Theorem for a linear time-discrete regulator problem with a quadratic performance index. Some properties were also investigated of the Kalman filter serving the process state estimation. An analysis is reported for a stochastic case, this for both time-invariant and time-variant Kalman filter gain matrix. (author)

  2. Studies on the closed-loop digital control of multi-modular reactors

    International Nuclear Information System (INIS)

    Bernard, J.A.; Henry, A.F.; Lanning, D.D.; Meyer, J.E.

    1992-11-01

    This report describes the theoretical development and the evaluation via both experiment and simulation of digital methods for the closed-loop control of power, temperature, and steam generator level in multi-modular reactors. The major conclusion of the research reported here is that the technology is currently available to automate many aspects of the operation of multi-modular plants. This will in turn minimize the number of required personnel and thus contain both operating and personnel costs, allow each module to be operated at a different power level thereby staggering the times at which refuelings would be needed, and maintain the competitiveness of US industry relative to foreign vendors who are developing and applying advanced control concepts. The technology described in this report is appropriate to the proposed multi-modular reactor designs and to present-generation pressurized water reactors. Its extension to boiling water reactors is possible provided that the commitment is made to create a real-time model of a BWR. The work reported here was performed by the Massachusetts Institute of Technology (MIT) under contract to the Oak Ridge National Laboratory (ORNL) and to the United States Department of Energy (Division of Industry and University Programs, Contract No. DE-FG07-90ER12930.)

  3. Studies on the closed-loop digital control of multi-modular reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, J.A. [Massachusetts Inst. of Tech., Cambridge, MA (United States). Nuclear Reactor Lab.; Henry, A.F.; Lanning, D.D.; Meyer, J.E. [Massachusetts Inst. of Tech., Cambridge, MA (United States). Dept. of Nuclear Engineering

    1992-11-01

    This report describes the theoretical development and the evaluation via both experiment and simulation of digital methods for the closed-loop control of power, temperature, and steam generator level in multi-modular reactors. The major conclusion of the research reported here is that the technology is currently available to automate many aspects of the operation of multi-modular plants. This will in turn minimize the number of required personnel and thus contain both operating and personnel costs, allow each module to be operated at a different power level thereby staggering the times at which refuelings would be needed, and maintain the competitiveness of US industry relative to foreign vendors who are developing and applying advanced control concepts. The technology described in this report is appropriate to the proposed multi-modular reactor designs and to present-generation pressurized water reactors. Its extension to boiling water reactors is possible provided that the commitment is made to create a real-time model of a BWR. The work reported here was performed by the Massachusetts Institute of Technology (MIT) under contract to the Oak Ridge National Laboratory (ORNL) and to the United States Department of Energy (Division of Industry and University Programs, Contract No. DE-FG07-90ER12930.)

  4. A review of the MIT experiments on the closed-loop digital control of reactor power

    International Nuclear Information System (INIS)

    Bernard, J.A.

    1989-01-01

    In this paper a review is provided of certain major experiments conducted from 1985 to 1988 as part of the MIT program on the development and demonstration of advanced technologies for the closed-loop digital control of nuclear reactors. Included are demonstrations of the supervisory control of neutronic power using an alternate formulation of the dynamic period equation, the use of the MIT-SNL Period-Generated Minimum Time Control Laws for the time-optimal control of neutronic power, and the evaluation of predictive displays as an operator aid. The significance of each of these advances is discussed in terms of the overall development of a multi-tiered controller that includes supervisory algorithms, predictive control laws, and automated reasoning

  5. Design an optimal controller for nuclear reactor using a digital computer

    International Nuclear Information System (INIS)

    Saleh, F.M.A.

    1986-01-01

    An attempt is carried out to design an optimal controller, for a model nuclear reactor at one hand, and a model nuclear power plant at another hand using a digital computer. The design philosophy adopted was to specify the system dynamics in terms of a desired system transfer function, and realizing the design synthesis through state-variable feedback technique, thus ensuring both stability and optimization in the state space sense. The control design was also tested by carrying out digital simulation transient response runs (step, ramp, impulse, etc.) and agreement between the predicted desirable response and actual response of the overall design was achieved. Furthermore the performance of the controller is verified against a reference non-linear model for purposes of assessing the accuracy of the linearized approximation model. The results show that state-variable feedback policy can rank as an effective optimal technique for designing control algorithm for an on-line computer of a nuclear power plant. 41 figs. 43 refs

  6. Cyber security level assignment for research reactor digital instrumentation and control system architecture using concept of defense in depth

    International Nuclear Information System (INIS)

    Shin, Jin Soo; Heo, Gyun Young; Son, Han Seong; Kim, Young Ki; Park, Jaek Wan

    2012-01-01

    Due to recent aging of the analog instrumentation of many nuclear power plants (NPPs) and research reactors, the system reliability decreases while maintenance and testing costs increase. In addition, it is difficult to find the substitutable analog equipment s due to obsolescence. Therefore, the instrumentation and control (I and C) systems have changed from analog system to digital system due to these facts. With the introduction of digital systems, research reactors are forced to care for the problem of cyber attacks because I and C systems have been digitalized using networks or communication systems. Especially, it is more issued at research reactors due to the accessibility of human resources. In the real world, an IBM researcher has been successful in controlling the software by penetrating a NPPs network in U.S. on July 2008 and acquiring the control right of nuclear facilities after one week. Moreover, the malignant code called 'stuxnet' impaired the nearly 1,000 centrifugal separators in Iran according to an IAEA report. The problem of cyber attacks highlights the important of cyber security, which should be emphasized. Defense.in.depth (DID) is a significant concept for the cyber security to work properly. DID institutes and maintains a hardy program for critical digital asset (CDA) by implementing multiple security boundaries. In this work, we assign cyber security levels to a typical digital I and C system using DID concept. This work is very useful in applying the concept of DID to nuclear industry with respect to cyber security

  7. New digital control system for the operation of the Colombian research reactor IAN-R1; Nuevo sistema de control digital para la operacion del reactor de investigacion Colombiano IAN-R1

    Energy Technology Data Exchange (ETDEWEB)

    Celis del A, L.; Rivero, T.; Bucio, F.; Ramirez, R.; Segovia, A.; Palacios, J., E-mail: lina.celis@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    En 2011, Mexico won the Colombian international tender for the renewal of instrumentation and control of the IAN-R1 Reactor, to Argentina and the United States. This paper presents the design criteria and the development made for the new digital control system installed in the Colombian nuclear reactor IAN-R1, which is based on a redundant and diverse architecture, which provides increased availability, reliability and safety in the reactor operation. This control system and associated instrumentation met all national export requirements, with the safety requirements established by the IAEA as well as the requirements demanded by the Colombian Regulatory Body in nuclear matter. On August 20, 2012, the Colombian IAN-R1 reactor reached its first criticality controlled with the new system developed at Instituto Nacional de Investigaciones Nucleares (ININ). On September 14, 2012, the new control system of the Colombian IAN-R1 reactor was officially handed over to the Colombian authorities, this being the first time that Mexico exported nuclear technology through the ININ. Currently the reactor is operating successfully with the new control system, and has an operating license for 5 years. (Author)

  8. Technology Transfer Programme In Reactor Digital Instrumentation And Control System (REDICS) Project: Knowledge, Experiences And Future Expectations

    International Nuclear Information System (INIS)

    Nurfarhana Ayuni Joha; Mohamad Puad Abu; Izhar Abu Hussin; Ridzuan Abdul Mutalib; Zareen Khan Abdul Jalil Khan; Mohd Khairulezwan Abdul Manan; Mohd Sabri Minhat; Mohd Idris Taib

    2013-01-01

    The PUSPATI TRIGA MARK II research reactor in Malaysia was commissioned in 1982. After 31 years of operation, Nuclear Malaysia is taking an approach for a better research and development in nuclear radiations as well as the technical services that provided. Reactor TRIGA PUSPATI (RTP) is currently upgrading its control console from analogue to digital system. The Reactor Digital Instrumentation and Control System (ReDICS) project is done on cooperation with Korea Atomic Energy Research Institute (KAERI), Korea including the technical part from the design stage until commissioning as well as the Technology Transfer Program (TTP). TTP in this ReDICS project is a part of Human Resource and System Development Program. It was carried out from the design stage until the commissioning of the system. It covers all subjects related to the design on the digital system and the requirements for the operation of RTP. The objective of this paper is to share the knowledge and experiences gained through this ReDICS project. This paper will also discuss the future expectations from this ReDICS project for Nuclear Malaysia and its personnel, as well as to the country. (author)

  9. Design of a decoupled AP1000 reactor core control system using digital proportional–integral–derivative (PID) control based on a quasi-diagonal recurrent neural network (QDRNN)

    Energy Technology Data Exchange (ETDEWEB)

    Wei, Xinyu, E-mail: xyuwei@mail.xjtu.edu.cn; Wang, Pengfei, E-mail: pengfeixiaoli@yahoo.cn; Zhao, Fuyu, E-mail: fuyuzhao_xj@163.com

    2016-08-01

    Highlights: • We establish a disperse dynamic model for AP1000 reactor core. • A digital PID control based on QDRNN is used to design a decoupling control system. • The decoupling performance is verified and discussed. • The decoupling control system is simulated under the load following operation. - Abstract: The control system of the AP1000 reactor core uses the mechanical shim (MSHIM) strategy, which includes a power control subsystem and an axial power distribution control subsystem. To address the strong coupling between the two subsystems, an interlock between the two subsystems is used, which can only alleviate but not eliminate the coupling. Therefore, sometimes the axial offset (AO) cannot be controlled tightly, and the flexibility of load-following operation is limited. Thus, the decoupling of the original AP1000 reactor core control system is the focus of this paper. First, a two-node disperse dynamic model is established for the AP1000 reactor core to use PID control. Then, a digital PID control system based on a quasi-diagonal recurrent neural network (QDRNN) is designed to decouple the original system. Finally, the decoupling of the control system is verified by the step signal and load-following condition. The results show that the designed control system can decouple the original system as expected and the AO can be controlled much more tightly. Moreover, the flexibility of the load following is increased.

  10. Digital driver of alternate current motors of the control rods in a nuclear research reactor

    International Nuclear Information System (INIS)

    Sainz M, E.

    1996-01-01

    The updating of the instruments as the operation console of the TRIGA Mark III Salazar Reactor is based on the use of a personal computer that works as data acquisition and control device. The power changes on the reactor have been made through the inserting or extraction of four control rods, that they are operated by mechanisms based in alternate current motors. That is with the object to handling each of the bars and so avoiding too the degradation about the performance of the computer of process. Also it is using four drives of smart kind which do the basic duties for generating the control signals and verifying the sensors state of the limits in continuous form. The computer and drivers are organized as a ring net using the serial port R S-232. The computer of process sends the orders and the identification of destination instrument throughout the net. (Author)

  11. Digital computer operation of a nuclear reactor

    Science.gov (United States)

    Colley, R.W.

    1982-06-29

    A method is described for the safe operation of a complex system such as a nuclear reactor using a digital computer. The computer is supplied with a data base containing a list of the safe state of the reactor and a list of operating instructions for achieving a safe state when the actual state of the reactor does not correspond to a listed safe state, the computer selects operating instructions to return the reactor to a safe state.

  12. Direct digital control of furnaces irradiated in nuclear reactors; Surveillance et regulation multiplexee par calculateur numerique de fours irradies

    Energy Technology Data Exchange (ETDEWEB)

    Joumard, R. [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1969-07-01

    An experimental direct digital control system has been realised in the 'C.E.N.G.', in order to verify that a computer makes easier the control of the experiments done in the nuclear reactors and to solve the theoretical and technical difficulties. The regulation is applied to thermal processes. The sampled data systems theory permits to choose the type of an efficient and simple digital compensator, and to establish a diagram which gives the values of the correcting parameters (obtained by minimizing the difference between the output and the input when perturbations occur). The programme execute, in simultaneity, supervision and regulation. Complex possibilities of printing out measures and alarms existed. The computer works out an incremental correction which makes step motors to turn. These motors act on the heating organs. The theoretical values and answers have been confirmed. The accuracy was limited essentially by the input quantification (1/1000 th). The comfort of such a system has been noticeable. (author) [French] Une installation de controle numerique direct fut realisee a titre experimental au C.E.N.G pour verifier qu'un ordinateur rendait plus aisee l'exploitation des experiences faites en pile nucleaire et pour degager les difficultes theoriques et techniques. La regulation s'applique a des processus thermiques. La theorie des systemes echantillonnes a permis de choisir un type de correcteur numerique simple et efficace et d'etablir un abaque qui donne les valeurs des parametres correcteurs minimisant les ecarts enregistres entre la reponse et la consigne en presence de perturbations. Le programme effectuait simultanement de la surveillance et de la regulation. Une restitution complexe des informations et des alarmes sur machine a ecrire etait possible. Le calculateur elaborait une correction incrementielle qui faisait tourner des moteurs pas a pas, lesquels commandaient les organes de puissance de chauffage. Les valeurs

  13. Reactor control device

    International Nuclear Information System (INIS)

    Fukami, Haruo; Morimoto, Yoshinori.

    1981-01-01

    Purpose: To operate a reactor always with safety operation while eliminating the danger of tripping. Constitution: In a reactor control device adapted to detect the process variants of a reactor, control a control rod drive controlling system based on the detected signal to thereby control the driving the control rods, control the reactor power and control the electric power generated from an electric generator by the output from the reactor, detection means is provided for the detection of the electric power from said electric generator, and a compensation device is provided for outputting control rod driving compensation signals to the control rod driving controlling system in accordance with the amount of variation in the detected value. (Seki, T.)

  14. Reactor power control device

    International Nuclear Information System (INIS)

    Ishii, Yoshihiko; Arita, Setsuo; Miyamoto, Yoshiyuki; Fukazawa, Yukihisa; Ishii, Kazuhiko

    1998-01-01

    The present invention provides a reactor power control device capable of enhancing an operation efficiency while keeping high reliability and safety in a BWR type nuclear power plant. Namely, the device of the present invention comprises (1) a means for inputting a set value of a generator power and a set value of a reactor power, (2) a means for controlling the reactor power to either smaller one of the reactor power corresponding to the set value of the generator power and the set value of the reactor power. With such procedures, even if the nuclear power plant is set so as to operate it to make the reactor power 100%, when the generator power reaches the upper limit, the reactor power is controlled with a preference given to the upper limit value of the generator power. Accordingly, safety and reliability are not deteriorated. The operation efficiency of the plant can be improved. (I.S.)

  15. Reactor power control device

    International Nuclear Information System (INIS)

    Kobayashi, Akira.

    1980-01-01

    Purpose: To prevent misoperation in a control system for the adjustment of core coolant flow rate, and the increase in the neutron flux density caused from the misoperation in BWR type reactors. Constitution: In a reactor power control system adapted to control the reactor power by the adjustment of core flow rate, average neutron flux signals of a reactor core, entire core flow rate signals and operation state signals for coolant recycling system are inputted to a microcomputer. The outputs from the computer are sent to a recycling MG set speed controller to control the reactor core flow rate. The computer calculates the change ratio with time in the average neutron flux signals, correlation between the average neutron flux signals and the entire core flow rate signals, change ratio with time in the operation state signals for the coolant recycling system and the like and judges the abnormality in the coolant recycling system based on the calculated results. (Ikeda, J.)

  16. Reactor power control device

    International Nuclear Information System (INIS)

    Imaruoka, Hiromitsu.

    1994-01-01

    A high pressure water injection recycling system comprising injection pipelines of a high pressure water injection system and a flow rate control means in communication with a pool of a pressure control chamber is disposed to a feedwater system of a BWR type reactor. In addition, the flow rate control means is controlled by a power control device comprising a scram impossible transient event judging section, a required injection flow rate calculation section for high pressure water injection system and a control signal calculation section. Feed water flow rate to be supplied to the reactor is controlled upon occurrence of a scram impossible transient event of the reactor. The scram impossible transient event is judged based on reactor output signals and scram operation demand signals and injection flow rate is calculated based on a predetermined reactor water level, and condensate storage tank water or pressure control chamber pool water is injected to the reactor. With such procedures, water level can be ensured and power can be suppressed. Further, condensate storage tank water of low enthalpy is introduced to the pressure suppression chamber pool to directly control elevation of water temperature and ensure integrity of the pressure vessel and the reactor container. (N.H.)

  17. Nuclear reactor control column

    International Nuclear Information System (INIS)

    Bachovchin, D.M.

    1982-01-01

    The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest crosssectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor

  18. Reactor power control system

    International Nuclear Information System (INIS)

    Tomisawa, Teruaki.

    1981-01-01

    Purpose: To restore reactor-power condition in a minimum time after a termination of turbine bypass by reducing the throttling of the reactor power at the time of load-failure as low as possible. Constitution: The transient change of the internal pressure of condenser is continuously monitored. When a turbine is bypassed, a speed-control-command signal for a coolant recirculating pump is generated according as the internal pressure of the condenser. When the signal relating to the internal pressure of the condenser indicates insufficient power, a reactor-control-rod-drive signal is generated. (J.P.N.)

  19. Development of Digital MMIS for Research Reactors: Graded Approaches

    Energy Technology Data Exchange (ETDEWEB)

    Khalil ur, Rahman; Shin, Jin Soo; Heo, Gyun Young [Kyunghee University, Yongin (Korea, Republic of); Son, Han Seong [Joongbu University, Geumsan (Korea, Republic of); Kim, Young Ki; Park, Jae Kwan; Seo, Sang Mun; Kim, Yong Jun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    Though research reactors are small in size yet they are important in terms of industrial applications and R and D, educational purposes. Keeping the eye on its importance, Korean government has intention to upgrade and extend this industry. Presently, Korea is operating only HANARO at Korea Atomic Energy Research Institute (KAERI) and AGN-201K at Kyung Hee University (KHU), which are not sufficient to meet the current requirements of research and education. In addition, we need self-sufficiency in design and selfreliance in design and operation, as we are installing research reactors in domestic as well as foreign territories for instance Jordan. Based on these demands, KAERI and universities initiated a 5 year research project since December 2011 collaboratly, for the deep study of reactor core, thermal hydraulics, materials and instrumentation and control (I and C). This particular study is being carried out to develop highly reliable advanced digital I and C systems using a grading approach. It is worth mentioning that next generation research reactor should be equipped with advance state of the art digital I and C for safe and reliable operation and impermeable cyber security system that is needed to be devised. Moreover, human error is one of important area which should be linked with I and C in terms of Man Machine Interface System (MMIS) and development of I and C should cover human factor engineering. Presently, the digital I and C and MMIS are well developed for commercial power stations whereas such level of development does not exist for research reactors in Korea. Since the functional and safety requirements of research reactors are not so strict as commercial power plants, the design of digital I and C systems for research reactors seems to be graded based on the stringency of regulatory requirements. This paper was motivated for the introduction of those missions, so it is going to describe the general overview of digital I and C systems, the graded

  20. Development of Digital MMIS for Research Reactors: Graded Approaches

    International Nuclear Information System (INIS)

    Khalil ur, Rahman; Shin, Jin Soo; Heo, Gyun Young; Son, Han Seong; Kim, Young Ki; Park, Jae Kwan; Seo, Sang Mun; Kim, Yong Jun

    2012-01-01

    Though research reactors are small in size yet they are important in terms of industrial applications and R and D, educational purposes. Keeping the eye on its importance, Korean government has intention to upgrade and extend this industry. Presently, Korea is operating only HANARO at Korea Atomic Energy Research Institute (KAERI) and AGN-201K at Kyung Hee University (KHU), which are not sufficient to meet the current requirements of research and education. In addition, we need self-sufficiency in design and selfreliance in design and operation, as we are installing research reactors in domestic as well as foreign territories for instance Jordan. Based on these demands, KAERI and universities initiated a 5 year research project since December 2011 collaboratly, for the deep study of reactor core, thermal hydraulics, materials and instrumentation and control (I and C). This particular study is being carried out to develop highly reliable advanced digital I and C systems using a grading approach. It is worth mentioning that next generation research reactor should be equipped with advance state of the art digital I and C for safe and reliable operation and impermeable cyber security system that is needed to be devised. Moreover, human error is one of important area which should be linked with I and C in terms of Man Machine Interface System (MMIS) and development of I and C should cover human factor engineering. Presently, the digital I and C and MMIS are well developed for commercial power stations whereas such level of development does not exist for research reactors in Korea. Since the functional and safety requirements of research reactors are not so strict as commercial power plants, the design of digital I and C systems for research reactors seems to be graded based on the stringency of regulatory requirements. This paper was motivated for the introduction of those missions, so it is going to describe the general overview of digital I and C systems, the graded

  1. Reactor control device

    International Nuclear Information System (INIS)

    Araki, Takao; Inoue, Toyokazu.

    1981-01-01

    Purpose: To protect the reactor floor by alleviating the shock imparted to the reactor floor by a dropped control rod when a wire rope accidentally breaks. Constitution: A control rod is hung by wire rope from a control rod drive, and shock absorbers are mounted at the upper and lower portions of the control rod. The outer diameter of the upper shock absorber is made larger than the inner diameter of a control rod inserting hole formed in the reactor core. If the control rod drops, the upper absorber is stopped at the upper tapered portion of the inserting hole. Thus, the dropping energy of the control rod can be sufficiently absorbed by the upper and lower shock absorbers. (Kamimura, M.)

  2. Reactor control system. PWR

    International Nuclear Information System (INIS)

    2009-01-01

    At present, 23 units of PWR type reactors have been operated in Japan since the start of Mihama Unit 1 operation in 1970 and various improvements have been made to upgrade operability of power stations as well as reliability and safety of power plants. As the share of nuclear power increases, further improvements of operating performance such as load following capability will be requested for power stations with more reliable and safer operation. This article outlined the reactor control system of PWR type reactors and described the control performance of power plants realized with those systems. The PWR control system is characterized that the turbine power is automatic or manually controlled with request of the electric power system and then the nuclear power is followingly controlled with the change of core reactivity. The system mainly consists of reactor automatic control system (control rod control system), pressurizer pressure control system, pressurizer water level control system, steam generator water level control system and turbine bypass control system. (T. Tanaka)

  3. KS-150 reactor control

    International Nuclear Information System (INIS)

    Wagner, K.

    1974-01-01

    A thorough description is presented of the control and protection system of the Bohunice A-1 reactor. The system including auxiliary facilities was developed, manufactured and installed at the reactor by the SKODA Works, Plzen. The system parameters are listed and a brief account is also given of the development efforts and of the physical and power start-up of the A-1 nuclear power plant. (L.O.)

  4. Reactor core control device

    International Nuclear Information System (INIS)

    Sano, Hiroki

    1998-01-01

    The present invention provides a reactor core control device, in which switching from a manual operation to an automatic operation, and the control for the parameter of an automatic operation device are facilitated. Namely, the hysteresis of the control for the operation parameter by an manual operation input means is stored. The hysteresis of the control for the operation parameter is collected. The state of the reactor core simulated by an operation control to which the collected operation parameters are manually inputted is determined as an input of the reactor core state to the automatic input means. The record of operation upon manual operation is stored as a hysteresis of control for the operation parameter, but the hysteresis information is not only the result of manual operation of the operation parameter. This is results of operation conducted by a skilled operator who judge the state of the reactor core to be optimum. Accordingly, it involves information relevant to the reactor core state. Then, it is considered that the optimum automatic operation is not deviated greatly from the manual operation. (I.S.)

  5. Reactor power control device

    International Nuclear Information System (INIS)

    Nishiyama, Hiroyuki.

    1991-01-01

    The device of the present invention prevents unnecessary automatic reactor shutdown, without increasing operator's burden by automatic insertion of selected control rods in case if a recycling pump in a BWR reaction should stop. That is, the device of the present invention comprises (1) a means for detecting that at least one recycling pump stops, (2) a means for judging region for inserting the selected control rods based on the reactor power and the recycling flowrate of driving water, and (3) a means for calculating a logic product of output signals sent from both of the means described above and outputting a selected control rod insertion signal. With such a constitution, if at least one recycling pump stops, the means (1) detects it. Further, the means (2) judges the regions for inserting the selected control rods. Then, the means (3) outputs a signal for inserting the selected control rods. As a result, since a group of control rods selected previously are inserted into the reactor rapidly, the reactor power is suppressed, to avoid the automatic reactor shutdown. (I.N.)

  6. Digital study of nuclear reactor instrument

    International Nuclear Information System (INIS)

    Lv Gongxiang; Yang Zhijun

    2006-01-01

    The paper introduces the design method of nuclear reactor's digital instrument developed by authors based on the AT89C52 single chip microcomputer. Also the instrument system hardware structure and software framework are given. The instrument apply DDC112 which is responsible for the measure of lower current. When designing the instrument system, anti-interference measure of software, especially hardware is considered seriously. (authors)

  7. Reactor instrumentation and control

    International Nuclear Information System (INIS)

    Wach, D.; Beraha, D.

    1980-01-01

    The methods for measuring radiation are shortly reviewed. The instrumentation for neutron flux measurement is classified into out-of-core and in-core instrumentation. The out-of-core instrumentation monitors the operational range from the subcritical reactor to full power. This large range is covered by several measurement channels which derive their signals from counter tubes and ionization chambers. The in-core instrumentation provides more detailed information on the power distribution in the core. The self-powered neutron detectors and the aeroball system in PWR reactors are discussed. Temperature and pressure measurement devices are briefly discussed. The different methods for leak detection are described. In concluding the plant instrumentation part some new monitoring systems and analysis methods are presented: early failure detection methods by noise analysis, acoustic monitoring and vibration monitoring. The presentation of the control starts from an qualitative assessment of the reactor dynamics. The chosen control strategy leads to the definition of the part-load diagram, which provides the set-points for the different control systems. The tasks and the functions of these control systems are described. In additiion to the control, a number of limiting systems is employed to keep the reactor in a safe operating region. Finally, an outlook is given on future developments in control, concerning mainly the increased application of process computers. (orig./RW)

  8. TREAT Reactor Control and Protection System

    International Nuclear Information System (INIS)

    Lipinski, W.C.; Brookshier, W.K.; Burrows, D.R.; Lenkszus, F.R.; McDowell, W.P.

    1985-01-01

    The main control algorithm of the Transient Reactor Test Facility (TREAT) Automatic Reactor Control System (ARCS) resides in Read Only Memory (ROM) and only experiment specific parameters are input via keyboard entry. Prior to executing an experiment, the software and hardware of the control computer is tested by a closed loop real-time simulation. Two computers with parallel processing are used for the reactor simulation and another computer is used for simulation of the control rod system. A monitor computer, used as a redundant diverse reactor protection channel, uses more conservative setpoints and reduces challenges to the Reactor Trip System (RTS). The RTS consists of triplicated hardwired channels with one out of three logic. The RTS is automatically tested by a digital Dedicated Microprocessor Tester (DMT) prior to the execution of an experiment. 6 refs., 5 figs., 1 tab

  9. A digital instrument for reactivity measurements in a nuclear reactor

    International Nuclear Information System (INIS)

    Chwaszczewski, S.

    1979-01-01

    An instrument for digital determination of the reactivity in nuclear reactors is described. It is based on the CAMAC standard apparatus, suitable for the use of pulse or current type neutron detectors and operates with prompt response and an output signal proportional to the core neutron flux. The measured data of neutron flux and reactivity can be registered by a digital display unit, an indicator, or, by request of the operator, a paper type punch. The algorithms used for reactivity calculation are considered and the results of numerical studies on those algorithms are discussed. The instrument has been used for determining the reactivity of the control elements in the fast-thermal assembly ANNA and in the research reactor MARIA. Some results of these measurements are given. (author)

  10. Nuclear reactor operation control process

    International Nuclear Information System (INIS)

    Doi, T.; Hiranuma, H.; Nishida, C.; Suematsu, S.

    1981-01-01

    A process for controlling operation of a nuclear reactor is described in which first control means is operated to cause reactor power to rise to a level at which a pellet-clad-mechanical-interaction begins to take place between a cladding and pellets of a fuel element. After interrupting the operation of the first control means, second control means is operated to cause the reactor power to rise to a preset level, the second control means being capable of effecting finer control of the reactor power than the first control means. When the reactor power deviates from the preset level with the progress of the reactor operation in the preset level, the second control means is operated so as to maintain the reactor power at the preset level

  11. Reactor water level control device

    International Nuclear Information System (INIS)

    Utagawa, Kazuyuki.

    1993-01-01

    A device of the present invention can effectively control fluctuation of a reactor water level upon power change by reactor core flow rate control operation. That is, (1) a feedback control section calculates a feedwater flow rate control amount based on a deviation between a set value of a reactor water level and a reactor water level signal. (2) a feed forward control section forecasts steam flow rate change based on a reactor core flow rate signal or a signal determining the reactor core flow rate, to calculate a feedwater flow rate control amount which off sets the steam flow rate change. Then, the sum of the output signal from the process (1) and the output signal from the process (2) is determined as a final feedwater flow rate control signal. With such procedures, it is possible to forecast the steam flow rate change accompanying the reactor core flow rate control operation, thereby enabling to conduct preceding feedwater flow rate control operation which off sets the reactor water level fluctuation based on the steam flow rate change. Further, a reactor water level deviated from the forecast can be controlled by feedback control. Accordingly, reactor water level fluctuation upon power exchange due to the reactor core flow rate control operation can rapidly be suppressed. (I.S.)

  12. Digital control rod blocking monitor

    International Nuclear Information System (INIS)

    Funayama, Yoshio.

    1996-01-01

    The present invention system is used for monitoring of a power region of a reactor, and used for monitoring of simultaneous withdrawal of a plurality of control rods without increasing the size or complicating the system. Namely, the system processes signals from a neutron flux detectors at the periphery of control rods controlled for withdrawal. As a result of the processing, the digital monitoring system generates an alarm when the reactor power at the periphery of the control rods fluctuates exceeding an allowable range. In the system, a control rod information forming means prepares frame data comprising front data, positions of the control rods to be withdrawn, frame numbers and completion data. A serial data transmitting means transmits the frame data successively as repeating frame data rows. A control rod information receiving means takes up the frame data of each of control rods to be withdrawn from the transmitted frame data rows. Since the system of the present invention can monitor the withdrawal of a plurality of control rods simultaneously without increasing the size or complicating the system, cost can be saved and the maintenance can be improved. (I.S.)

  13. Digital-image processing improves man-machine communication at a nuclear reactor

    International Nuclear Information System (INIS)

    Cook, S.A.; Harrington, T.P.; Toffer, H.

    1982-01-01

    The application of digital image processing to improve man-machine communication in a nuclear reactor control room is illustrated. At the Hanford N Reactor, operated by UNC Nuclear Industries for the United States Department of Energy, in Richland, Washington, digital image processing is applied to flow, temperature, and tube power data. Color displays are used to present the data in a clear and concise fashion. Specific examples are used to demonstrate the capabilities and benefits of digital image processing of reactor data. N Reactor flow and power maps for routine reactor operations and for perturbed reactor conditions are displayed. The advantages of difference mapping are demonstrated. Image processing techniques have also been applied to results of analytical reactor models; two examples are shown. The potential of combining experimental and analytical information with digital image processing to produce predictive and adaptive reactor core models is discussed. The applications demonstrate that digital image processing can provide new more effective ways for control room personnel to assess reactor status, to locate problems and explore corrective actions. 10 figures

  14. Programmable Digital Controller

    Science.gov (United States)

    Wassick, Gregory J.

    2012-01-01

    An existing three-channel analog servo loop controller has been redesigned for piezoelectric-transducer-based (PZT-based) etalon control applications to a digital servo loop controller. This change offers several improvements over the previous analog controller, including software control over proportional-integral-derivative (PID) parameters, inclusion of other data of interest such as temperature and pressure in the control laws, improved ability to compensate for PZT hysteresis and mechanical mount fluctuations, ability to provide pre-programmed scanning and stepping routines, improved user interface, expanded data acquisition, and reduced size, weight, and power.

  15. Experimental development of power reactor intelligent control

    International Nuclear Information System (INIS)

    Edwards, R.M.; Garcia, H.E.; Lee, K.Y.

    1992-01-01

    The US nuclear utility industry initiated an ambitious program to modernize the control systems at a minimum of ten existing nuclear power plants by the year 2000. That program addresses urgent needs to replace obsolete instrumentation and analog controls with highly reliable state-of-the-art computer-based digital systems. Large increases in functionality that could theoretically be achieved in a distributed digital control system are not an initial priority in the industry program but could be logically considered in later phases. This paper discusses the initial development of an experimental sequence for developing, testing, and verifying intelligent fault-accommodating control for commercial nuclear power plant application. The sequence includes an ultra-safe university research reactor (TRIGA) and a passively safe experimental power plant (Experimental Breeder Reactor 2)

  16. Nuclear reactor kinetics and control

    International Nuclear Information System (INIS)

    Lewins, J.

    1978-01-01

    A consistent, integrated account of modern developments in the study of nuclear reactor kinetics and the problem of their efficient and safe control. It aims to prepare the student for advanced study and research or practical work in the field. Special features include treatments of noise theory, reliability theory and safety related studies. It covers all aspects of the operation and control of nuclear reactors, power and research and is complete in providing physical data methods of calculation and solution including questions of equipment reliability. The work uses illustrations of the main types of reactors in use in the UK, USA and Europe. Each chapter contains problems and worked examples suitable for course work and study. The subject is covered in chapters, entitled: introductory review; neutron and precursor equations; elementary solutions at low power; linear reactor process dynamics with feedback; power reactor control systems; fluctuations and reactor noise; safety and reliability; nonlinear systems (safety and control); analogue computing. (author)

  17. Reactor power control device

    International Nuclear Information System (INIS)

    Watanabe, Mitsutaka

    1997-01-01

    Hardware of an analog nuclear instrumentation system is reformed, a function generator is added to a setting calculation circuit of the nuclear instrumentation system, and each of setting lines of the nuclear instrumentation system is set in parallel with an upper limit curve in an operation region defined by a second order or third order equation. Upon transient change of abnormal power elevation during operation, scram signals are generated by power change in the same state as 100% rated operation due to elevation of reactor thermal power. Since the operation limit value relative to transient change due to power elevation can be made substantially equal with the same as that upon rated operation, the operation limit value for partial power operation state can be kept substantially the same level as that upon rated operation. When transition change caused by abnormal control rod withdrawal occurs during operation, a control rod withdrawal inhibition signal can ensure the power elevation width equal with that upon rated power operation, and since the withdrawal inhibition signal is generated in substantially the same withdrawing state, the operation limit value relative to a partial power operation state can be kept at the same level as that during rated operation. (N.H.)

  18. Reactor water level control device

    International Nuclear Information System (INIS)

    Hiramatsu, Yohei.

    1980-01-01

    Purpose: To increase the rapid response of the waterlevel control converting a reactor water level signal into a non-linear type, when the water level is near to a set value, to stabilize the water level reducting correlatively the reactor water level variation signal to stabilize greatly from the set value, and increasing the variation signal. Constitution: A main vapor flow quality transmitter detects the vapor flow generated in a reactor and introduced into a turbine. A feed water flow transmitter detects the quantity of a feed water flow from the turbine to the reactor, this detected value is sent to an addition operating apparatus. On the other hand, the power signal of the reactor water level transmitter is sent to the addition operating apparatus through a non-linear water level signal converter. The addition operation apparatus generates a signal for requesting the feed water flow quantity from both signals. Upon this occasion, the reactor water level signal converter makes small the reactor water level variation when the reactor level is close the set value, and when the water level deviates greatly from the set value, the reactor water level variation is made large thereby to improve the rapid response of the reactor coater level control. (Yoshino, Y.)

  19. Nuclear reactor with control rods

    International Nuclear Information System (INIS)

    Obermeyer, F.D.; Berringer, R.T.

    1979-01-01

    A liquid-cooled nuclear reactor including fuel assemblies mounted within a reactor vessel having linearly movable control rods passing through control rod guide tubes into respective aligned fuel assemblies is described. Reactor coolant circulates through the assemblies. Guide tubes and other vessel internals structures located above the assemblies and is discharged through an outlet nozzle positioned above the elevation of primary flow openings in the guide tube walls. The guide tube includes internal horizontal supports and a length limited continuous control rod guide which, in conjunction with the flow openings, alleviate detrimental coolant cross flows and frictional restraints imposed upon the control rods

  20. Safety aspect of digital reactor protection system in Japan

    International Nuclear Information System (INIS)

    Ogiso, Zen-Ichi

    1998-01-01

    It was early in 1980's that the digital controllers were first applied to nuclear power plant in japan. After that, their application area had been expanding gradually, reaching to the overall integrated digital system including the safety system in Kashiwazaki-Kariwa units 6 and 7. The software for computer-based systems has been produced using the graphical language ''POL'' in Japanese nuclear power plants. It is the fundamental principle that the reliability of the software should be assured through the properly managed quality assurance. The POL-based system is fitted to this principle. In applying POL-based systems to safety system, the MITI, Ministry of International Trade and Industry, identified the licensing issues as the regulatory body, while the utilities had developed the digital technology feasible to the safety application. Through the activities, a specific industrial design guide for the software important to safety was established and the adequacy of the technology was certified through the demonstration tests of the integrated system. In the safety examination of the digital reactor protection system of K-6/7, the application of POL were approved. The POL-based systems in nuclear power plants were successful design and production process of the POL-based systems. This paper describes the activities in licensing and maintaining the computer-based systems by the utilities and manufacturers as well as the MITI. (author)

  1. Nuclear reactor control rod

    International Nuclear Information System (INIS)

    Cearley, J.E.; Izzo, K.R.

    1987-01-01

    This patent describes a vertically oriented bottom entry control rod from a nuclear reactor: a frame including an elongated central spine of cruciform cross section connected between an upper support member and a lower support member both of cruciform shape having four laterally extending arms. The arms are in alignment with the arms of the lower support member and each aligned upper and lower support members has a sheath extending between; absorber plates of neutron absorber material, different from the material of the frame, one of the absorber plates is positioned within a sheath beneath each of the arms; attachment means suspends the absorber plates from the arms of the upper support member within a sheath; elongated absorber members positioned within a sheath between each of the suspended absorber plates and an arm of the lower support member; and joint means between the upper ends of the absorber members and the lower ends of the suspended absorber plates for minimizing gaps; the sheath means encloses the suspended absorber plates and the absorber members extending between aligned arms of the upper and lower support members and secured

  2. Light Water Reactor Sustainability Program. Digital Architecture Requirements

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Kenneth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Oxstrand, Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    The Digital Architecture effort is a part of the Department of Energy (DOE) sponsored Light-Water Reactor Sustainability (LWRS) Program conducted at Idaho National Laboratory (INL). The LWRS program is performed in close collaboration with industry research and development (R&D) programs that provides the technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants (NPPs). One of the primary missions of the LWRS program is to help the U.S. nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. Therefore, a major objective of the LWRS program is the development of a seamless digital environment for plant operations and support by integrating information from plant systems with plant processes for nuclear workers through an array of interconnected technologies. In order to get the most benefits of the advanced technology suggested by the different research activities in the LWRS program, the nuclear utilities need a digital architecture in place to support the technology. A digital architecture can be defined as a collection of information technology (IT) capabilities needed to support and integrate a wide-spectrum of real-time digital capabilities for nuclear power plant performance improvements. It is not hard to imagine that many processes within the plant can be largely improved from both a system and human performance perspective by utilizing a plant wide (or near plant wide) wireless network. For example, a plant wide wireless network allows for real time plant status information to easily be accessed in the control room, field workers’ computer-based procedures can be updated based on the real time plant status, and status on ongoing procedures can be incorporated into smart schedules in the outage command center to allow for more accurate planning of critical tasks. The goal

  3. Reactor control rod supporting structure

    International Nuclear Information System (INIS)

    Akimoto, Tokuzo; Miyata, Hiroshi.

    1984-01-01

    Purpose: To enable stable reactor core control even in extremely great vertical earthquakes, as well as under normal operation conditions in FBR type reactors. Constitution: Since a mechanism for converting the rotational movement of a control rod into vertical movement is placed at the upper portion of the reactor core at high temperature, the mechanism should cause fusion or like other danger after the elapse of a long period of time. In view of the above, the conversion mechanism is disposed to the lower portion of the reactor core at a lower temperature region. Further, the connection between the control rod and the control rod drive can be separated upon great vertical earthquakes. (Seki, T.)

  4. The resonance absorption controlled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Caro, R.

    1977-07-01

    In this report a new method of reactor control based on tho isotopic moderator composition variation is studied, taking as a reference a D{sub 2}O/H{sub 2}O system. With this method an spectacular increment in the burn-up degree and a sensible reduction of the conventional control system is obtained. An important part of this work has been the detailed analysis of the parameters affecting the neutron spectrum in a heterogeneous reactor. (Author) 50 refs.

  5. The resonance absorption controlled reactor

    International Nuclear Information System (INIS)

    Caro, R.

    1977-01-01

    In this report a new method of reactor control based on tho isotopic moderator composition variation is studied, taking as a reference a D 2 O/H 2 O system. With this method an spectacular increment in the burn-up degree and a sensible reduction of the conventional control system is obtained. An important part of this work has been the detailed analysis of the parameters affecting the neutron spectrum in a heterogeneous reactor. (Author) 50 refs

  6. Control rod drive of nuclear reactor

    International Nuclear Information System (INIS)

    Zhuchkov, I.I.; Gorjunov, V.S.; Zaitsev, B.I.

    1980-01-01

    This invention relates to nuclear reactors and, more particularly, to a drive of a control rod of a nuclear reactor and allows power control, excess reactivity compensation, and emergency shut-down of a reactor. (author)

  7. Evolution of reactor control modes

    International Nuclear Information System (INIS)

    Mourlevat, J.L.

    2007-01-01

    The article reviews the different reactor control modes: mode A, mode G and mode X that are used in PWR reactors designed by Areva. The purpose of reactor controlling is to compensate reactivity effects (xenon poisoning and counter-reaction effects) generated by load changes. A control mode is the strategy followed by using both soluble boron and the control rods to handle these reactivity effects. Soluble boron plays an important role in mode A but is less efficiency toward the end-of-cycle. Generally soluble boron is used to compensate slow reactivity effects. Mode G is based on the optimization of the use of the control rods and allows a quick return to the nominal power. Mode X combines the uses of control rods and soluble boron to cope with the operator's wishes: sparing effluents or having a quick return to nominal power. The mode X appears to be more penalizing concerning the fuel-cladding interaction. (A.C.)

  8. REACTOR CONTROL ROD OPERATING SYSTEM

    Science.gov (United States)

    Miller, G.

    1961-12-12

    A nuclear reactor control rod mechanism is designed which mechanically moves the control rods into and out of the core under normal conditions but rapidly forces the control rods into the core by catapultic action in the event of an emergency. (AEC)

  9. Technique of nuclear reactors controls

    International Nuclear Information System (INIS)

    Weill, J.

    1953-12-01

    This report deal about 'Techniques of control of the nuclear reactors' in the goal to achieve the control of natural uranium reactors and especially the one of Saclay. This work is mainly about the measurement into nuclear parameters and go further in the measurement of thermodynamic variables,etc... putting in relief the new features required on behalf of the detectors because of their use in the thermal neutrons flux. In the domain of nuclear measurement, we indicate the realizations and the results obtained with thermal neutron detectors and for the measurement of ionizations currents. We also treat the technical problem of the start-up of a reactor and of the reactivity measurement. We give the necessary details for the comprehension of all essential diagrams and plans put on, in particular, for the reactor of Saclay. (author) [fr

  10. Method of controlling reactor operation

    International Nuclear Information System (INIS)

    Masuda, Hiroyuki.

    1982-01-01

    Purpose: To prevent fuel failures, as well as enable easy control of power fluctuation due to transient changes in the xenon density. Method: Upon actuation of a control valve, heavy water containing poisons flows through a poison removing tower and the poisons are removed by chemical resins charged in the removing tower. The heavy water flows passing through the heavy water inlet pipe into a reactor core. As the result, neutron absorption in the reactor core is decreased to increase the reactor power. Then, neutron fluxes in the reactor core are detected and the reactor power from a power converter is compared with the output from a power-up ratio setter in a power judging device to control a helium control valve to thereby decrease or increase the heavy water level. While on the other hand, the output from an operation signal generator is sent to a memory unit for the heavy water level operation time and the control time for the liquid poison density is corrected based on the control time for the moderator liquid level, whereby fuel soundness can be maintained. (Yoshino, Y.)

  11. Computerized reactor monitor and control for research reactors

    International Nuclear Information System (INIS)

    Buerger, L.; Vegh, E.

    1981-09-01

    The computerized process control system developed in the Central Research Institute for Physics, Budapest, Hungary, is described together with its special applications at research reactors. The nuclear power of the Hungarian research reactor is controlled by this computerized system, too, while in Lybia many interesting reactor-hpysical calculations are built into the computerized monitor system. (author)

  12. Nuclear reactor power control device

    International Nuclear Information System (INIS)

    Koshi, Yuji; Sakata, Akira; Karatsu, Hiroyuki.

    1987-01-01

    Purpose: To control abrupt changes in neutron fluxes by feeding back a correction signal obtained from a deviation between neutron fluxes and heat fluxes for changing the reactor core flow rate to a recycling flow rate control system upon abrupt power change of a nuclear reactor. Constitution: In addition to important systems, that is, a reactor pressure control system and a recycling control system in the power control device of a BWR type power plant, a control circuit for feeding back a deviation between neutron fluxes and heat fluxes to a recycling flow rate control system is disposed. In the suppression circuit, a deviation signal is prepared in an adder from neutron flux and heat flux signals obtained through a primary delay filter. The deviation signal is passed through a dead band and an advance/delay filter into a correction signal, which is adapted to be fed back to the recycling flow rate control system. As a result, the reactor power control can be conducted smoothly and it is possible to effectively suppress the abrupt change or over shoot of the neutron fluxes and abrupt power change. (Kamimura, M.)

  13. Design of controller for control rod of research reactors

    International Nuclear Information System (INIS)

    Abou-Zaid, R.M.F.M

    2008-01-01

    Designing and testing digital control system for any nuclear research reactor can be costly and time consuming. In this thesis, a rapid, low-cost proto typing and testing procedure for digital controller design is proposed using the concept of Hardware-In-The-Loop (HIL). Some of the control loop components are real hardware components and the others are simulated. First, the whole system is modeled and tested by Real-Time Simulation (RTS) using conventional simulation techniques such as MATLAB / SIMULINK. Second the Hardware-in-the-loop simulation is tested using Real-Time Windows Target in MATLAB and Visual C ++ . The control parts are included as hardware components which are the reactor control rod and its drivers. Three kinds of controllers are studied, Proportional-Derivative (PD), Proportional-Integral-Derivative (PID) and Fuzzy controller. An experimental setup for the hardware used in HIL concept for the control of the nuclear research reactor has been realized. Experimental results are obtained and compared with the simulation results. The experimental results indicate the validation of HIL method in this domain.

  14. Application of fault tree methodology to modeling of the AP1000 plant digital reactor protection system

    International Nuclear Information System (INIS)

    Teolis, D.S.; Zarewczynski, S.A.; Detar, H.L.

    2012-01-01

    The reactor trip system (RTS) and engineered safety features actuation system (ESFAS) in nuclear power plants utilizes instrumentation and control (IC) to provide automatic protection against unsafe and improper reactor operation during steady-state and transient power operations. During normal operating conditions, various plant parameters are continuously monitored to assure that the plant is operating in a safe state. In response to deviations of these parameters from pre-determined set points, the protection system will initiate actions required to maintain the reactor in a safe state. These actions may include shutting down the reactor by opening the reactor trip breakers and actuation of safety equipment based on the situation. The RTS and ESFAS are represented in probabilistic risk assessments (PRAs) to reflect the impact of their contribution to core damage frequency (CDF). The reactor protection systems (RPS) in existing nuclear power plants are generally analog based and there is general consensus within the PRA community on fault tree modeling of these systems. In new plants, such as AP1000 plant, the RPS is based on digital technology. Digital systems are more complex combinations of hardware components and software. This combination of complex hardware and software can result in the presence of faults and failure modes unique to a digital RPS. The United States Nuclear Regulatory Commission (NRC) is currently performing research on the development of probabilistic models for digital systems for inclusion in PRAs; however, no consensus methodology exists at this time. Westinghouse is currently updating the AP1000 plant PRA to support initial operation of plants currently under construction in the United States. The digital RPS is modeled using fault tree methodology similar to that used for analog based systems. This paper presents high level descriptions of a typical analog based RPS and of the AP1000 plant digital RPS. Application of current fault

  15. Nuclear reactor power control device

    International Nuclear Information System (INIS)

    Takigawa, Yukio; Ebata, Shigeo.

    1991-01-01

    Possibility for the occurrence of vibrations in a reactor power due to lowering of reactor core stability is annihilated by avoiding an operation near natural convection high power state of a BWR type reactor. That is, a deviation between a total pump speed signal sent from a total pump speed calculation device and a total pump speed demand signal sent from a recycling flow rate control system is calculated in a deviation calculation device, and it is inputted to a comparison device. When the deviation is greater than a predetermined value, the comparison device judges it as a trip of the recycling pump, and outputs an actuation signal to a selection control rod insertion device to insert a predetermined number of control rods. As a result, the output at the natural convection state is decreased to lower than that of a 80% power flow rate control curve. Further, when the deviation value is smaller than the predetermined value, an actuation signal is outputted to the recycling pump speed controller so that the pump speed is not decreased to lower than a lowest pump speed. As a result, the lower limit of the reactor core flow rate is limited to the flow rate corresponding to the lowest pump speed. (I.S.)

  16. A computer control system for a research reactor

    International Nuclear Information System (INIS)

    Crawford, K.C.; Sandquist, G.M.

    1987-01-01

    Most reactor applications until now, have not required computer control of core output. Commercial reactors are generally operated at a constant power output to provide baseline power. However, if commercial reactor cores are to become load following over a wide range, then centralized digital computer control is required to make the entire facility respond as a single unit to continual changes in power demand. Navy and research reactors are much smaller and simpler and are operated at constant power levels as required, without concern for the number of operators required to operate the facility. For navy reactors, centralized digital computer control may provide space savings and reduced personnel requirements. Computer control offers research reactors versatility to efficiently change a system to develop new ideas. The operation of any reactor facility would be enhanced by a controller that does not panic and is continually monitoring all facility parameters. Eventually very sophisticated computer control systems may be developed which will sense operational problems, diagnose the problem, and depending on the severity of the problem, immediately activate safety systems or consult with operators before taking action

  17. A digital data acquisition and display system for ITU TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Can, B.; Omuz, S.

    2008-01-01

    Full text: In this study, a digital data acquisition and display system realized for ITU TRIGA Mark-II Reactor is described. This system is realized in order to help the reactor operator and to increase reactor console capacity. The system consists of two main units, which are host computers and RTI-815F, analog devices, data acquisition card. RTI-815F is multi-function analog/digital input/output board that plugs into one of the available long expansion slots in the IBM-PC, PC/XT, PC/AT, or equivalent personal computers. It has 16 analog input channels for single-ended input signals or 8 analog input channels for differential input signals. But its channel capacity can be increased to 32 input channels for single-ended input signals or 16 input channels for differential input signals. RTI-815F board contains 2 analog output channels, 8 digital input channels and 8 digital output channels. In the ITD TRIGA Mark-II Reactor, 6 fuel temperature channels, 3 water temperature channels, 3 control rod position channels and 4 power channels are chosen as analog input signals for RTI-815F. Its digital outputs are assigned to cooling tower fan, primary and secondary pump reactor scram, control rod rundown. During operation, data are automatically archived to disk and displayed on screen. The channel selection time and sampling time can be adjusted. The simulated movement and position of control rods in the reactor core can be noted and displayed. The changes of power, fuel temperature and water temperature can be displayed on the screen as a graphic. In this system both period and reactivity are calculated and displayed on the screen. (authors)

  18. Reactor shutdown control device

    International Nuclear Information System (INIS)

    Itooka, Satoshi; Matsushima, Shusuke; Otsuki, Jun.

    1988-01-01

    Purpose: To obtain a device capable of attaining stable temperature for sodium around a magnetic body and capable of adequately changing the temperature at which control rods are fallen even after the mounting to the device. Constitution: One of fuel assemblies situated around control rod assemblies is selected by selective introduction pipe of fuel assemblies, thereby introducing sodium always at stable temperature from the selected assembly to the periphery of the magnetic body under the normal state and prevent the misoperation under the normal operation by receiving only the sodium temperature. Further, by changing the current supply state to the solenoid magnet for setting the temperature at which the control rod falls upon abnormality, it is possible to adequately fall the control rods. (Yoshihara, H.)

  19. Control rod for nuclear reactor

    International Nuclear Information System (INIS)

    Tada, Kaoru; Kawano, Shohei

    1998-01-01

    A guide roller is prepared by forming an oxide membrane on the surface of a molded roller product comprising, as a material, a deposition-reinforced type nickel-based alloy reinforced by deposition of fine particles by applying a heat treatment to a nickel-based alloy. When the guide roller is used in reactor water, since the roller has an oxide membrane on the surface, leaching of nickel to reactor water is reduced, and radioactive corrosive products including cobalt 58 are reduced to decrease an operator's exposure dose upon periodical inspections of a plant. The oxide membrane is formed by applying heat treatment under an oxidative atmosphere. Then, the amount of abrasion of pins and rollers in association with start-up or shut down of a reactor and control of the power can be reduced thereby enabling to suppress increase of radiation dose due to cobalt 60 and cobalt 58. (N.H.)

  20. Twenty-fifth water reactor safety information meeting: Proceedings. Volume 3: Thermal hydraulic research and codes; Digital instrumentation and control; Structural performance

    International Nuclear Information System (INIS)

    Monteleone, S.

    1998-04-01

    This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following: (1) thermal hydraulic research and codes; (2) digital instrumentation and control; (3) structural performance

  1. ADAPTIVE CONTROL SYSTEM OF INDUSTRIAL REACTORS

    Directory of Open Access Journals (Sweden)

    Vyacheslav K. Mayevski

    2014-01-01

    Full Text Available This paper describes a mathematical model of an industrial chemical reactor for production of synthetic rubber. During reactor operation the model parameters vary considerably. To create a control algorithm performed transformation of mathematical model of the reactor in order to obtain a dependency that can be used to determine the model parameters are changing during reactor operation.

  2. Reactivity control assembly for nuclear reactor

    Science.gov (United States)

    Bollinger, Lawrence R.

    1984-01-01

    Reactivity control assembly for nuclear reactor comprises supports stacked above reactor core for holding control rods. Couplers associated with the supports and a vertically movable drive shaft have lugs at their lower ends for engagement with the supports.

  3. Controlled thermonuclear fusion reactors

    International Nuclear Information System (INIS)

    Walstrom, P.L.

    1976-01-01

    Controlled production of energy by fusion of light nuclei has been the goal of a large portion of the physics community since the 1950's. In order for a fusion reaction to take place, the fuel must be heated to a temperature of 100 million degrees Celsius. At this temperature, matter can exist only in the form of an almost fully ionized plasma. In order for the reaction to produce net power, the product of the density and energy confinement time must exceed a minimum value of 10 20 sec m -3 , the so-called Lawson criterion. Basically, two approaches are being taken to meet this criterion: inertial confinement and magnetic confinement. Inertial confinement is the basis of the laser fusion approach; a fuel pellet is imploded by intense laser beams from all sides and ignites. Magnetic confinement devices, which exist in a variety of geometries, rely upon electromagnetic forces on the charged particles of the plasma to keep the hot plasma from expanding. Of these devices, the most encouraging results have been achieved with a class of devices known as tokamaks. Recent successes with these devices have given plasma physicists confidence that scientific feasibility will be demonstrated in the next generation of tokamaks; however, an even larger effort will be required to make fusion power commercially feasible. As a result, emphasis in the controlled thermonuclear research program is beginning to shift from plasma physics to a new branch of nuclear engineering which can be called fusion engineering, in which instrumentation and control engineers will play a major role. Among the new problem areas they will deal with are plasma diagnostics and superconducting coil instrumentation

  4. Controlled thermonuclear fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Walstrom, P.L.

    1976-01-01

    Controlled production of energy by fusion of light nuclei has been the goal of a large portion of the physics community since the 1950's. In order for a fusion reaction to take place, the fuel must be heated to a temperature of 100 million degrees Celsius. At this temperature, matter can exist only in the form of an almost fully ionized plasma. In order for the reaction to produce net power, the product of the density and energy confinement time must exceed a minimum value of 10/sup 20/ sec m/sup -3/, the so-called Lawson criterion. Basically, two approaches are being taken to meet this criterion: inertial confinement and magnetic confinement. Inertial confinement is the basis of the laser fusion approach; a fuel pellet is imploded by intense laser beams from all sides and ignites. Magnetic confinement devices, which exist in a variety of geometries, rely upon electromagnetic forces on the charged particles of the plasma to keep the hot plasma from expanding. Of these devices, the most encouraging results have been achieved with a class of devices known as tokamaks. Recent successes with these devices have given plasma physicists confidence that scientific feasibility will be demonstrated in the next generation of tokamaks; however, an even larger effort will be required to make fusion power commercially feasible. As a result, emphasis in the controlled thermonuclear research program is beginning to shift from plasma physics to a new branch of nuclear engineering which can be called fusion engineering, in which instrumentation and control engineers will play a major role. Among the new problem areas they will deal with are plasma diagnostics and superconducting coil instrumentation.

  5. Safety evaluation of the KNICS digital reactor protection system

    International Nuclear Information System (INIS)

    Kang, Hyun Gook; Jang, Seung Cheol; Choi, Jong Gyun

    2005-01-01

    Korean Nuclear I and C System (KNICS) project, a national research program for developing a safety-class digital system, has designed a new reactor protection system (RPS). The usage of digital equipment in a safety critical application increases the importance of a risk evaluation since microprocessors and software technologies make the digital system very complex and their unavailability is hard to quantify. This paper addresses the safety evaluation of the KNICS RPS in consideration of the several technical concerns of a safety modeling for a digital system. We also present the fault-tree modeling technique and the risk evaluation results. A fault-tree model which includes the common cause failure events, the coverage of a fault-tolerant mechanism and the software failure event is developed. Based on the minimal cut sets of the model, we discuss the system unavailability of the newly developed design of the KNICS RPS

  6. Digital System Reliability Test for the Evaluation of safety Critical Software of Digital Reactor Protection System

    Directory of Open Access Journals (Sweden)

    Hyun-Kook Shin

    2006-08-01

    Full Text Available A new Digital Reactor Protection System (DRPS based on VME bus Single Board Computer has been developed by KOPEC to prevent software Common Mode Failure(CMF inside digital system. The new DRPS has been proved to be an effective digital safety system to prevent CMF by Defense-in-Depth and Diversity (DID&D analysis. However, for practical use in Nuclear Power Plants, the performance test and the reliability test are essential for the digital system qualification. In this study, a single channel of DRPS prototype has been manufactured for the evaluation of DRPS capabilities. The integrated functional tests are performed and the system reliability is analyzed and tested. The results of reliability test show that the application software of DRPS has a very high reliability compared with the analog reactor protection systems.

  7. Digitally controlled distributed phase shifter

    Science.gov (United States)

    Hietala, Vincent M.; Kravitz, Stanley H.; Vawter, Gregory A.

    1993-01-01

    A digitally controlled distributed phase shifter is comprised of N phase shifters. Digital control is achieved by using N binary length-weighted electrodes located on the top surface of a waveguide. A control terminal is attached to each electrode thereby allowing the application of a control signal. The control signal is either one or two discrete bias voltages. The application of the discrete bias voltages changes the modal index of a portion of the waveguide that corresponds to a length of the electrode to which the bias voltage is applied, thereby causing the phase to change through the underlying portion of the waveguide. The digitally controlled distributed phase shift network has a total phase shift comprised of the sum of the individual phase shifters.

  8. Spectral shift reactor control method

    International Nuclear Information System (INIS)

    Impink, A.J. Jr.

    1981-01-01

    A method of operating a nuclear reactor having a core and coolant displacer elements arranged in the core wherein is established a reator coolant temperature set point at which it is desired to operate said reactor and first reactor coolant temperature band limits are provided within which said set point is located and it is desired to operate said reactor charactrized in that said reactor coolant displacer elements are moved relative to the reactor core for adjusting the volume of reactor coolant in said core as said reactor coolant temperature approaches said first band limits thereby to maintain said reactor coolant temperature near said set point and within said first band limits

  9. Computerized reactor power regulation with logarithmic controller

    International Nuclear Information System (INIS)

    Gossanyi, A.; Vegh, E.

    1982-11-01

    A computerized reactor control system has been operating at a 5 MW WWR-SM research reactor in the Central Research Institute for Physics, Budapest, for some years. This paper describes the power controller used in the SPC operating mode of the system, which operates in a 5-decade wide power range with +-0.5% accuracy. The structure of the controller easily limits the minimal reactor period and produces a reactor transient with constant period if the power demand changes. (author)

  10. Architectural conceptual definition of the CAREM-25 reactor's control system

    International Nuclear Information System (INIS)

    Perez, J.C.; Santome, D.; Drexler, J.; Escudero, S.

    1990-01-01

    This work presents the conceptual definition of the CAREM 25 reactor's digital and monitoring control system structure. The requirements of the system are analyzed and different implementation alternatives are studied where possible basic architectures of the system and its topology are considered and evaluated. (Author) [es

  11. Bus-oriented digital control techniques in nuclear power plants

    International Nuclear Information System (INIS)

    Salm, M.

    1987-01-01

    The author states the conservative principles which govern the authorization procedures for nuclear reactor control systems. Using the example of a feedwater supply regulator, employing a digital, bus-oriented control system, he describes how the stigma attached to the word nuclear can be alleviated. (G.T.H.)

  12. Backfitting in Rossendorf research reactor control and instrumentation system

    International Nuclear Information System (INIS)

    Klebau, J.; Seidler, S.

    1985-01-01

    The paper generally describes a decentralized Hierarchical Information System (HIS) which has been developed for backfitting in Rossendorf Research Reactor (RFR) control and instrumentation system. The RFR was put into operation in 1957 and reconstructed from 2 MW up to a thermal power of 10 MW at the end of the sixties. Backfitting is planned by use of an advanced computerized control system for the next years. Main tasks of HIS are: Processmonitoring, online-disturbance analysis, technical diagnosis, direct digital control and use of a special industrial robot for discharging of irradiated materials out of the reactor. Experiences obtained by HIS during a testperiod will be presented. (author)

  13. Propose Reactor Control and Monitoring System for RTP

    International Nuclear Information System (INIS)

    Mohd Sabri Minhat; Izhar Abu Hussin; Mohd Idris Taib; Mohd Khairulezwan Abdul Manan; Nurfarhana Ayuni Joha

    2011-01-01

    Reactor control and monitoring system is a one of the important features used in reactor. The control and monitoring must come together to provide safety, excellent performance and reliable in nuclear reactor technology application. Objectives of this technical paper are to design and propose reactor control system and reactor monitoring system in Research Reactor (RTP) for Reactor Upgrading Project. (author)

  14. Power control system in BWR type reactors

    International Nuclear Information System (INIS)

    Nishizawa, Yasuo.

    1980-01-01

    Purpose: To control the reactor power so that the power distribution can satisfy the limiting conditions, by regulating the reactor core flow rate while monitoring the power distribution in the reactor core of a BWR type reactor. Constitution: A power distribution monitor determines the power distribution for the entire reactor core based on the data for neutron flux, reactor core thermal power, reactor core flow rate and control rod pattern from the reactor and calculates the linear power density distribution. A power up ratio computing device computes the current linear power density increase ratio. An aimed power up ratio is determined by converting the electrical power up ratio transferred from a load demand input device into the reactor core thermal power up ratio. The present reactor core thermal power up ratio is subtracted from the limiting power up ratio and the difference is sent to an operation amount indicator and the reactor core flow rate is changed in a reactor core flow rate regulator, by which the reactor power is controlled. (Moriyama, K.)

  15. Digital control in power electronics

    CERN Document Server

    Buso, Simone

    2015-01-01

    This book presents the reader, whether an electrical engineering student in power electronics or a design engineer, a selection of power converter control problems and their basic digital solutions, based on the most widespread digital control techniques. The presentation is primarily focused on different applications of the same power converter topology, the half-bridge voltage source inverter, considered both in its single- and three-phase implementation. This is chosen as the test case because, besides being simple and well known, it allows the discussion of a significant spectrum of the mo

  16. Digital control in power electronics

    CERN Document Server

    Buso, Simone

    2006-01-01

    This book presents the reader, whether an electrical engineering student in power electronics or a design engineer, some typical power converter control problems and their basic digital solutions, based on the most widespread digital control techniques. The presentation is focused on different applications of the same power converter topology, the half-bridge voltage source inverter, considered both in its single- and three-phase implementation. This is chosen as the case study because, besides being simple and well known, it allows the discussion of a significant spectrum of the more frequent

  17. Advanced digital rod position indication system for existing and next generation nuclear reactors

    International Nuclear Information System (INIS)

    Hashemian, H.M.; Morton, G.W.; Shumaker, B.D.

    2009-01-01

    The designs of many existing pressurized water reactors (PWRs) incorporate a digital rod position indication (DRPI) system to monitor the positions of the control and shutdown rods within the reactor. Recently, aging and obsolescence issues have led to an increase in problems with the DRPI systems including analog card failures and coil cable connection problems. These problems, along with plans for plant life extension, have prompted the industry to actively seek new options to monitor the health and accuracy of these DRPI systems in order to ensure reliable plant operation. (author)

  18. Study on modeling technology in digital reactor system

    International Nuclear Information System (INIS)

    Liu Xiaoping; Luo Yuetong; Tong Lili

    2004-01-01

    Modeling is the kernel part of a digital reactor system. As an extensible platform for reactor conceptual design, it is very important to study modeling technology and develop some kind of tools to speed up preparation of all classical computing models. This paper introduces the background of the project and basic conception of digital reactor. MCAM is taken as an example for modeling and its related technologies used are given. It is an interface program for MCNP geometry model developed by FDS team (ASIPP and HUT), and designed to run on windows system. MCAM aims at utilizing CAD technology to facilitate creation of MCNP geometry model. There have been two ways for MCAM to utilize CAD technology: (1) Making use of user interface technology in aid of generation of MCNP geometry model; (2) Making use of existing 3D CAD model to accelerate creation of MCNP geometry model. This paper gives an overview of MCAM's major function. At last, several examples are given to demonstrate MCAM's various capabilities. (authors)

  19. Monitoring and control of anaerobic reactors

    DEFF Research Database (Denmark)

    Pind, Peter Frode; Angelidaki, Irini; Ahring, Birgitte Kiær

    2003-01-01

    The current status in monitoring and control of anaerobic reactors is reviewed. The influence of reactor design and waste composition on the possible monitoring and control schemes is examined. After defining the overall control structure, and possible control objectives, the possible process mea...... control approaches that have been used are comprehensively described. These include simple and adaptive controllers, as well as more recent developments such as fuzzy controllers, knowledge-based controllers and controllers based on neural networks....

  20. Monitoring and control of anaerobic reactors

    DEFF Research Database (Denmark)

    Pind, Peter Frode; Angelidaki, Irini; Ahring, Birgitte Kiær

    2003-01-01

    The current status in monitoring and control of anaerobic reactors is reviewed. The influence of reactor design and waste composition on the possible monitoring and control schemes is examined. After defining the overall control structure, and possible control objectives, the possible process...

  1. A digital method for period measurements in a nuclear reactor

    International Nuclear Information System (INIS)

    Mundim, Sergio Gorretta

    1971-02-01

    The present paper begins by giving a theoretical treatment for the nuclear reactor period. The conventional method of measuring the period is analysed and some previously developed digital methods are described. The paper criticises the latter, pointing out some deficiencies which the proposed process is able to eliminate. All errors connected with this process are also analysed. The paper presents suitable solutions to reduce them to a minimum. The total error is found to he less than the error presented by the other methods described. A digital period meter is designed with memory resources and an automatic scaler changer. Integrated circuits specifications are used in it. Real time experiments with nuclear reactors were made in order to check te validity of the method. The data acquired were applied to a simulated digital period meter implemented in a general purpose computer. The nuclear part of the work was developed at the 'Comissao Nacional de Energia Nuclear' and the simulation work was dane at the 'Departamento de Calculo Cientifico' of COPPE, which also advised the author in the completion of this thesis. (author)

  2. Safety regulations concerning instrumentation and control systems for research reactors

    International Nuclear Information System (INIS)

    El-Shanshoury, A.I.

    2009-01-01

    A brief study on the safety and reliability issues related to instrumentation and control systems in nuclear reactor plants is performed. In response, technical and strategic issues are used to accomplish instrumentation and control systems safety. For technical issues there are ; systems aspects of digital I and C technology, software quality assurance, common-mode software, failure potential, safety and reliability assessment methods, and human factors and human machine interfaces. The strategic issues are the case-by-case licensing process and the adequacy of the technical infrastructure. The purpose of this work was to review the reliability of the safety systems related to these technical issues for research reactors

  3. Autonomous Control of Space Nuclear Reactors Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Nuclear reactors to support future robotic and manned missions impose new and innovative technological requirements for their control and protection instrumentation....

  4. Updating reactor control: mini-computers

    International Nuclear Information System (INIS)

    Crawford, K.C.; Sandquist, G.M.

    1984-01-01

    An aging reactor control console and a limited operating budget have impeded many research projects in the TRIGA reactor facility at the University of Utah. The, University's present console is Circa 1959 vintage and repairs to the console are frequently required which present many electronic problems to a staff with little electronic training. As an alternative to a single function control console we are developing a TRIGA control system based upon a mini-computer. The system hardware has been specified and the hardware is currently being acquired. The software will be programmed by the staff to customize the system to the reactor's physical systems and technical specifications. The software will be designed to monitor and control all reactor functions, control a pneumatic sample transfer system, acquire and analyze neutron activation data, provide reactor facility security surveillance, provide reactor documentation including online logging of physical parameters, and record regularly scheduled reactor calibrations and laboratory accounting procedures. The problem of hardware rewiring and changing technical specifications and changing safety system characteristics can be easily handled in the software. Our TRIGA reactor also functions as a major educational resource using available reactor based software. The computer control system can be employed to provide on-line training in reactor physics and kinetics. (author)

  5. Investigation of Classification and Design Requirements for Digital Software for Advanced Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Park, Gee Young; Jung, H. S.; Ryu, J. S.; Park, C

    2005-06-15

    As the digital technology is being developed drastically, it is being applied to various industrial instrumentation and control (I and C) fields. In the nuclear power plants, I and C systems are also being installed by digital systems replacing their corresponding analog systems installed previously. There had been I and C systems constructed by analog technology especially for the reactor protection system in the research reactor HANARO. Parallel to the pace of the current trend for digital technology, it is desirable that all I and C systems including the safety critical and non-safety systems in an advanced research reactor is to be installed based on the computer based system. There are many attractable features in using digital systems against existing analog systems in that the digital system has a superior performance for a function and it is more flexible than the analog system. And any fruit gained from the newly developed digital technology can be easily incorporated into the existing digital system and hence, the performance improvement of a computer based system can be implemented conveniently and promptly. Moreover, the capability of high integrity in electronic circuits reduces the electronic components needed to construct the processing device and makes the electronic board simple, and this fact reveals that the hardware failure itself are unlikely to occur in the electronic device other than some electric problems. Balanced the fact mentioned above are the roles and related issues of the software loaded on the digital integrated hardware. Some defects in the course of software development might induce a severe damage on the computer system and plant systems and therefore it is obvious that comprehensive and deep considerations are to be placed on the development of the software in the design of I and C system for use in an advanced research reactor. The work investigates the domestic and international standards on the classifications of digital

  6. Experimental development of power reactor advanced controllers

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, R.M. (Pennsylvania State Univ., University Park, PA (United States). Dept. of Nuclear Engineering); Weng, C.K. (Pennsylvania State Univ., University Park, PA (United States). Dept. of Mechanical Engineering); Lindsay, R.W. (Argonne National Lab., Idaho Falls, ID (United States))

    1992-01-01

    A systematic approach for developing and verifying advanced controllers with potential application to commercial nuclear power plants is suggested. The central idea is to experimentally demonstrate an advanced control concept first on an ultra safe research reactor followed by demonstration on a passively safe experimental power reactor and then finally adopt the technique for improving safety, performance, reliability and operability at commercial facilities. Prior to completing an experimental sequence, the benefits and utility of candidate advanced controllers should be established through theoretical development and simulation testing. The applicability of a robust optimal observer-based state feedback controller design process for improving reactor temperature response for a TRIGA research reactor, Liquid Metal-cooled Reactor (LMR), and a commercial Pressurized Water Reactor (PWR) is presented to illustrate the potential of the proposed experimental development concept.

  7. Experimental development of power reactor advanced controllers

    Energy Technology Data Exchange (ETDEWEB)

    Edwards, R.M. [Pennsylvania State Univ., University Park, PA (United States). Dept. of Nuclear Engineering; Weng, C.K. [Pennsylvania State Univ., University Park, PA (United States). Dept. of Mechanical Engineering; Lindsay, R.W. [Argonne National Lab., Idaho Falls, ID (United States)

    1992-06-01

    A systematic approach for developing and verifying advanced controllers with potential application to commercial nuclear power plants is suggested. The central idea is to experimentally demonstrate an advanced control concept first on an ultra safe research reactor followed by demonstration on a passively safe experimental power reactor and then finally adopt the technique for improving safety, performance, reliability and operability at commercial facilities. Prior to completing an experimental sequence, the benefits and utility of candidate advanced controllers should be established through theoretical development and simulation testing. The applicability of a robust optimal observer-based state feedback controller design process for improving reactor temperature response for a TRIGA research reactor, Liquid Metal-cooled Reactor (LMR), and a commercial Pressurized Water Reactor (PWR) is presented to illustrate the potential of the proposed experimental development concept.

  8. Digital control programmer for temperature control

    International Nuclear Information System (INIS)

    Rajore, S.B.; Kumar, S.V.

    1993-01-01

    This report describes a PC based digital control programmer for controlling and programming temperature of a high vacuum resistance heating furnace and the software developed to control power using PID algorithm. It also describes the amplifier specially developed to suit the input requirement of the non-standard W5 thermocouple and the software and hardware protections introduced in the system. (author). 5 refs., 8 figs., 1 appendix

  9. A new control strategy for nuclear power reactors

    International Nuclear Information System (INIS)

    Wakabayashi, H.; Sasaki, K.; Takegaki, M.

    1990-01-01

    A new automatic direct digital control strategy for nuclear power reactors is presented. It is based on a simple control logic of comparison between the available time (the time for the error signal to disappear) and the required time (the time for the time derivative to match that of the target trend). The method aims to control the system to an acceptable state within a minimum time under a number of restraints. The control capability of the method is shown for two typical transients. This method is generally applicable to process control in which time-optimal control based on the maximum principle is sought

  10. Study, design and evaluation of nuclear reactor computer control system

    International Nuclear Information System (INIS)

    Menacer, S.

    1988-01-01

    Nuclear reactor control is a complex process that varies with each reactor and there is no universal agreement as to the best type of control system. After the use of conventional systems for a long time, attention turned towards digital techniques in the reactor control system. This interest emerged because of the difficulties faced in the data manipulation, mainly for post-incident analysis. However, it is not sufficient to insert a computer in a system to solve all the data-handling problems and also the insertion of a computer in a real-time system is not without any effect on the overall system. The scope of this thesis is to show the important parameters that have to be taken into account when choosing and evaluate the performances of the selected system

  11. Power control device for nuclear reactors

    International Nuclear Information System (INIS)

    Kagawa, Tatsuo

    1984-01-01

    Purpose: To eliminate for requirement of control rods and movable portions, as well as ensure the safety and reliability of the operation. Constitution: A plurality of control tubes are disposed within a reactor core instead of control rods. Tubes are connected from below the reactor core to the control tubes for supplying liquid poisons such as aqueous boric acid to the inside of the control tubes. Further, tubes are connected to the upper portion of the control tubes for guiding the liquid poisons from the reactor core to the outside. The tubes for supplying and discharging the liquid poisons are introduced externally through the flange disposed at the upper portion of a pressure vessel. At the outside of the pressure vessel, are disposed a liquid poison tank, a pressurizing source, a pressure control valve, a liquid level meter and the like. The control for the reactor power is conducted by controlling the level of the liquid poisons in the control tubes. (Ikeda, J.)

  12. NEUTRON DENSITY CONTROL IN A NEUTRONIC REACTOR

    Science.gov (United States)

    Young, G.J.

    1959-06-30

    The method and means for controlling the neutron density in a nuclear reactor is described. It describes the method and means for flattening the neutron density distribution curve across the reactor by spacing the absorbing control members to varying depths in the central region closer to the center than to the periphery of the active portion of the reactor to provide a smaller neutron reproduction ratio in the region wherein the members are inserted, than in the remainder of the reactor thereby increasing the over-all potential power output.

  13. Reactor core control device for FBR type reactor

    International Nuclear Information System (INIS)

    Iida, Norihiko

    1991-01-01

    The device of the present invention comprises a control line having a control pump and a control tank for injecting liquids for neutron reflectors to an annular tank disposed in a reactor container, a supply line having a supply pump and a supply tank for supplying the liquids for the reflectors to the control tank, a drain line having a control valve, a drain valve and a drain tank disposed to the annular tank and a make-up line for supplying the liquids for the reflectors from the drain tank to the control tank. Liquids such as water or oil for the neutron reflectors are injected in the annular tank disposed at the periphery of the reactor core to raise the level of the liquids in the tank and conduct burning in the reactor core. The liquid level may be controlled by an appropriate ON/OFF operation of a pump with no requirement for a motor or a driving device at a high accuracy and rotational portions. Periodical maintenances are not necessary. Reactor scram can be conducted only by opening the drain valve and the reflectors may be made of inexpensive materials. (N.H.)

  14. Reactor power control method and device

    International Nuclear Information System (INIS)

    Fushimi, Atsushi; Ishii, Yoshihiko; Miyamoto, Yoshiyuki; Ishii, Kazuhiko; Kiyoharu, Norihiko; Aizawa, Yuko.

    1997-01-01

    The present invention provides a method and a device suitable to rise the temperature and increase the pressure of the reactor to an aimed pressure in accordance with an aimed value for a reactor water temperature changing rate in the course of rising temperature and increasing pressure of the reactor upon start up of a BWR type power plant. Namely, neutron fluxes in the reactor and the temperature of reactor water are detected respectively. The maximum value among the detected values for the neutron fluxes is detected. The reactor water temperature changing rate is calculated based on the detected values of the reactor water temperature, from which the maximum value of the reactor water temperature changing rate is detected. An aimed value for the neutron flux is calculated in accordance with both detected maximum values and the aimed value of the reactor water temperature changing rate. The position of control rods is adjusted in accordance with the aimed value for the calculated neutron flux. Then, an aimed value for the neutron flux for realizing the aimed value for the reactor water temperature changing rate can be obtained accurately with no influence of the sensitivity of the detected values of the neutron fluxes and the time delay of the reactor water temperature changing rate. (I.S.)

  15. Parametric robustness of the reactor control system

    International Nuclear Information System (INIS)

    Lee, Yoon Joon; Kim, Sin

    1998-01-01

    The parametric analysis method is applied to the reactor control system. The mathematical reactor model is discussed in terms of the parametric uncertainties. The Tsyplin-Polyak locus, based on the boundary crossing theorem, shows the reactor plant has an intrinsic stability. A simple controller is incorporated to the plant to configure the overall closed loop system. Then parametric stability margins are obtained for the controller constants. The results show that a new design constraint of the controller robustness with respect to overall system should be considered for the controller design of an uncertain system

  16. Autonomous Control of Space Nuclear Reactors

    Science.gov (United States)

    Merk, John

    2013-01-01

    Nuclear reactors to support future robotic and manned missions impose new and innovative technological requirements for their control and protection instrumentation. Long-duration surface missions necessitate reliable autonomous operation, and manned missions impose added requirements for failsafe reactor protection. There is a need for an advanced instrumentation and control system for space-nuclear reactors that addresses both aspects of autonomous operation and safety. The Reactor Instrumentation and Control System (RICS) consists of two functionally independent systems: the Reactor Protection System (RPS) and the Supervision and Control System (SCS). Through these two systems, the RICS both supervises and controls a nuclear reactor during normal operational states, as well as monitors the operation of the reactor and, upon sensing a system anomaly, automatically takes the appropriate actions to prevent an unsafe or potentially unsafe condition from occurring. The RPS encompasses all electrical and mechanical devices and circuitry, from sensors to actuation device output terminals. The SCS contains a comprehensive data acquisition system to measure continuously different groups of variables consisting of primary measurement elements, transmitters, or conditioning modules. These reactor control variables can be categorized into two groups: those directly related to the behavior of the core (known as nuclear variables) and those related to secondary systems (known as process variables). Reliable closed-loop reactor control is achieved by processing the acquired variables and actuating the appropriate device drivers to maintain the reactor in a safe operating state. The SCS must prevent a deviation from the reactor nominal conditions by managing limitation functions in order to avoid RPS actions. The RICS has four identical redundancies that comply with physical separation, electrical isolation, and functional independence. This architecture complies with the

  17. Centralized digital control of accelerators

    International Nuclear Information System (INIS)

    Melen, R.E.

    1983-09-01

    In contrasting the title of this paper with a second paper to be presented at this conference entitled Distributed Digital Control of Accelerators, a potential reader might be led to believe that this paper will focus on systems whose computing intelligence is centered in one or more computers in a centralized location. Instead, this paper will describe the architectural evolution of SLAC's computer based accelerator control systems with respect to the distribution of their intelligence. However, the use of the word centralized in the title is appropriate because these systems are based on the use of centralized large and computationally powerful processors that are typically supported by networks of smaller distributed processors

  18. Fission control system for nuclear reactor

    Science.gov (United States)

    Conley, G.H.; Estes, G.P.

    Control system for nuclear reactor comprises a first set of reactivity modifying rods fixed in a reactor core with their upper ends stepped in height across the core, and a second set of reactivity modifying rods movable vertically within the reactor core and having their lower ends stepped to correspond with the stepped arrangement of the first set of rods, pairs of the rods of the first and second sets being in coaxial alignment.

  19. Reactivity control assembly for nuclear reactor. [LMFBR

    Science.gov (United States)

    Bollinger, L.R.

    1982-03-17

    This invention, which resulted from a contact with the United States Department of Energy, relates to a control mechanism for a nuclear reactor and, more particularly, to an assembly for selectively shifting different numbers of reactivity modifying rods into and out of the core of a nuclear reactor. It has been proposed heretofore to control the reactivity of a breeder reactor by varying the depth of insertion of control rods (e.g., rods containing a fertile material such as ThO/sub 2/) in the core of the reactor, thereby varying the amount of neutron-thermalizing coolant and the amount of neutron-capturing material in the core. This invention relates to a mechanism which can advantageously be used in this type of reactor control system.

  20. Low cost highly available digital control computer

    International Nuclear Information System (INIS)

    Silvers, M.W.

    1986-01-01

    When designing digital controllers for critical plant control it is important to provide several features. Among these are reliability, availability, maintainability, environmental protection, and low cost. An examination of several applications has lead to a design that can be produced for approximately $20,000 (1000 control points). This design is compatible with modern concepts in distributed and hierarchical control. The canonical controller element is a dual-redundant self-checking computer that communicates with a cross-strapped, electrically isolated input/output system. The input/output subsystem comprises multiple intelligent input/output cards. These cards accept commands from the primary processor which are validated, executed, and acknowledged. Each card may be hot replaced to facilitate sparing. The implementation of the dual-redundant computer architecture is discussed. Called the FS-86, this computer can be used for a variety of applications. It has most recently found application in the upgrade of San Francisco's Bay Area Rapid Transit (BART) train control currently in progress and has been proposed for feedwater control in a boiling water reactor

  1. Innovative Control concepts for German pressurized water reactors

    International Nuclear Information System (INIS)

    Brzozowski, Raphael; Kuhn, Andreas

    2010-01-01

    Controlling reactor power without any manual support is becoming more and more important. The READIG project (READIG = Reactor Instrumentation and Digital Control) power control system installed in unit 2 of the Philippsburg nuclear power station (KKP 2) requires no manual intervention except for specific strategy criteria settings. It was even possible to eliminate the power distribution set points. With minor adaptations, this concept can be applied in other PWR plants as well. KKP 2 is a PWR plant with particularly sophisticated core charges; as a consequence, the I and C systems were adapted accordingly. The increase in integral reactor power and the low-leakage core charges are the main reasons for lower limiting margins, especially in peak limiting. The standard control concept was supplemented in such a way that a more precise fine control concept for power distribution in the full-load regime is achieved. The READIG project fully utilizes the possibilities offered by digital TXS Technology, which is why use is also made of physical parameterization. The new power distribution control concept has these advantages: - Operation at small peak-/DNB-reactor output limitation margins. - Stable control without manual intervention also in load cycles and in the frequency control mode. - Simplified operation due to omission of the power distribution set point. - Reduction to zero of the frequency of L-bank steps at constant power with superimposed frequency control mode. - Reduction to zero of the frequency of D-bank steps at constant power with superimposed frequency control mode. - Lower quantities of demineralized water to be fed at constant power with superimposed frequency control mode (±1%). (orig.)

  2. ANALYTICAL SYNTHESIS OF CHEMICAL REACTOR CONTROL SYSTEM

    Directory of Open Access Journals (Sweden)

    Alexander Labutin

    2017-02-01

    Full Text Available The problem of the analytical synthesis of the synergetic control system of chemical reactor for the realization of a complex series-parallel exothermal reaction has been solved. The synthesis of control principles is performed using the analytical design method of aggregated regulators. Synthesized nonlinear control system solves the problem of stabilization of the concentration of target component at the exit of reactor and also enables one to automatically transfer to new production using the equipment.

  3. Distributed expert systems for nuclear reactor control

    International Nuclear Information System (INIS)

    Otaduy, P.J.

    1992-01-01

    A network of distributed expert systems is the heart of a prototype supervisory control architecture developed at the Oak Ridge National Laboratory (ORNL) for an advanced multimodular reactor. Eight expert systems encode knowledge on signal acquisition, diagnostics, safeguards, and control strategies in a hybrid rule-based, multiprocessing and object-oriented distributed computing environment. An interactive simulation of a power block consisting of three reactors and one turbine provides a realistic, testbed for performance analysis of the integrated control system in real-time. Implementation details and representative reactor transients are discussed

  4. ADVANCED CONTROL FOR A ETHYLENE REACTOR

    Directory of Open Access Journals (Sweden)

    Dumitru POPESCU

    2017-06-01

    Full Text Available The main objective of this work is the design and implementation of control solutions for petrochemical processes, namely the control and optimization of a pyrolysis reactor, the key-installation in the petrochemical industry. We present the technological characteristics of this petrochemical process and some aspects about the proposed control system solution for the ethylene plant. Finally, an optimal operating point for the reactor is found, considering that the process has a nonlinear multi-variable structure. The results have been implemented on an assembly of pyrolysis reactors on a petrochemical platform from Romania.

  5. The electronic temperature control and measurements reactor fuel rig circuits

    International Nuclear Information System (INIS)

    Glowacki, S.W.

    1980-01-01

    The electronic circuits of two digital temperature meters developed for the thermocouple of Ni-NiCr type are described. The output thermocouple signal as converted by means of voltage-to-freguency converter. The frequency is measured by a digital scaler controled by quartz generator signals. One of the described meter is coupled with digital temperature controler which drives the power stage of the reactor rig heater. The internal rig temperature is measured by the thermocouple providing the input signal to the mentioned voltage-to-frequency converter, that means the circuits work in the negative feedback loop. The converter frequency-to-voltage ratio is automatically adjusted to match to thermocouple sensitivity changes in the course of the temperature variations. The accuracy of measuring system is of order of +- 1degC for thermocouple temperature changes from 523 K up to 973 K (50degC up to 700degC). (author)

  6. GMA200 ATEGG Digital Controls Demonstration.

    Science.gov (United States)

    1979-10-01

    controlled on corrected speed by the digital controller through an Electro-Hydraulic Servo Valve ( EHSV ) and a position transducer. A pair of hydraulic...nozzle is hydraulically actuated. The digital controller effects position control via an EHSV and position transducers. e. Engine Sensors In addition...A-AG93 958 GENERAL MOTORS CORP INDIANAPOLIS IN DETROIT DIESEL A-ETC F/6 1/ r A - 3 5.S o79 GAM200 ATEGG DIGITAL CONTROLS DEMONSTRATION.(U) OCT 79 D

  7. Digital control card based on digital signal processor

    International Nuclear Information System (INIS)

    Hou Shigang; Yin Zhiguo; Xia Le

    2008-01-01

    A digital control card based on digital signal processor was developed. Two Freescale DSP-56303 processors were utilized to achieve 3 channels proportional- integral-differential regulations. The card offers high flexibility for 100 MeV cyclotron RF system development. It was used as feedback controller in low level radio frequency control prototype, with the feedback gain parameters continuously adjustable. By using high precision analog to digital converter with 500 kHz sampling rate, a regulation bandwidth of 20 kHz was achieved. (authors)

  8. Advanced Neutron Source reactor control and plant protection systems design

    International Nuclear Information System (INIS)

    Anderson, J.L.; Battle, R.E.; March-Leuba, J.; Khayat, M.I.

    1992-01-01

    This paper describes the reactor control and plant protection systems' conceptual design of the Advanced Neutron Source (ANS). The Plant Instrumentation, Control, and Data Systems and the Reactor Instrumentation and Control System of the ANS are planned as an integrated digital system with a hierarchical, distributed control structure of qualified redundant subsystems and a hybrid digital/analog protection system to achieve the necessary fast response for critical parameters. Data networks transfer information between systems for control, display, and recording. Protection is accomplished by the rapid insertion of negative reactivity with control rods or other reactivity mechanisms to shut down the fission process and reduce heat generation in the fuel. The shutdown system is designed for high functional reliability by use of conservative design features and a high degree of redundance and independence to guard against single failures. Two independent reactivity control systems of different design principles are provided, and each system has multiple independent rods or subsystems to provide appropriate margin for malfunctions such as stuck rods or other single failures. Each system is capable of maintaining the reactor in a cold shutdown condition independently of the functioning of the other system. A highly reliable, redundant channel control system is used not only to achieve high availability of the reactor, but also to reduce challenges to the protection system by maintaining important plant parameters within appropriate limits. The control system has a number of contingency features to maintain acceptable, off-normal conditions in spite of limited control or plant component failures thereby further reducing protection system challenges

  9. Reliability Analysis Study of Digital Reactor Protection System in Nuclear Power Plant

    International Nuclear Information System (INIS)

    Guo, Xiao Ming; Liu, Tao; Tong, Jie Juan; Zhao, Jun

    2011-01-01

    The Digital I and C systems are believed to improve a plants safety and reliability generally. The reliability analysis of digital I and C system has become one research hotspot. Traditional fault tree method is one of means to quantify the digital I and C system reliability. Review of advanced nuclear power plant AP1000 digital protection system evaluation makes clear both the fault tree application and analysis process to the digital system reliability. One typical digital protection system special for advanced reactor has been developed, which reliability evaluation is necessary for design demonstration. The typical digital protection system construction is introduced in the paper, and the process of FMEA and fault tree application to the digital protection system reliability evaluation are described. Reliability data and bypass logic modeling are two points giving special attention in the paper. Because the factors about time sequence and feedback not exist in reactor protection system obviously, the dynamic feature of digital system is not discussed

  10. Hardware replacements and software tools for digital control computers

    International Nuclear Information System (INIS)

    Walker, R.A.P.; Wang, B-C.; Fung, J.

    1996-01-01

    Technological obsolescence is an on-going challenge for all computer use. By design, and to some extent good fortune, AECL has had a good track record with respect to the march of obsolescence in CANDU digital control computer technology. Recognizing obsolescence as a fact of life, AECL has undertaken a program of supporting the digital control technology of existing CANDU plants. Other AECL groups are developing complete replacement systems for the digital control computers, and more advanced systems for the digital control computers of the future CANDU reactors. This paper presents the results of the efforts of AECL's DCC service support group to replace obsolete digital control computer and related components and to provide friendlier software technology related to the maintenance and use of digital control computers in CANDU. These efforts are expected to extend the current lifespan of existing digital control computers through their mandated life. This group applied two simple rules; the product, whether new or replacement should have a generic basis, and the products should be applicable to both existing CANDU plants and to 'repeat' plant designs built using current design guidelines. While some exceptions do apply, the rules have been met. The generic requirement dictates that the product should not be dependent on any brand technology, and should back-fit to and interface with any such technology which remains in the control design. The application requirement dictates that the product should have universal use and be user friendly to the greatest extent possible. Furthermore, both requirements were designed to anticipate user involvement, modifications and alternate user defined applications. The replacements for hardware components such as paper tape reader/punch, moving arm disk, contact scanner and Ramtek are discussed. The development of these hardware replacements coincide with the development of a gateway system for selected CANDU digital control

  11. Reactor core and control rod assembly in FBR type reactor

    International Nuclear Information System (INIS)

    Fujimura, Koji; Kawashima, Katsuyuki; Itooka, Satoshi.

    1993-01-01

    Fuel assemblies and control rod assemblies are attached respectively to reactor core support plates each in a cantilever fashion. Intermediate spacer pads are disposed to the lateral side of a wrapper tube just above the fuel rod region. Intermediate space pads are disposed to the lateral side of a control rod guide tube just above a fuel rod region. The thickness of the intermediate spacer pad for the control rod assembly is made smaller than the thickness of the intermediate spacer pad for the fuel assembly. This can prevent contact between intermediate spacer pads of the control guide tube and the fuel assembly even if the temperature of coolants is elevated to thermally expand the intermediate spacer pad, by which the radial displacement amount of the reactor core region along the direction of the height of the control guide tube is reduced substantially to zero. Accordingly, contribution of the control rod assembly to the radial expansion reactivity can be reduced to zero or negative level, by which the effect of the negative radial expansion reactivity of the reactor is increased to improve the safety upon thermal transient stage, for example, loss of coolant flow rate accident. (I.N.)

  12. Reactor feedwater pump control device

    International Nuclear Information System (INIS)

    Nishiyama, Hiroyuki.

    1990-01-01

    An amount of feedwater necessary for ensuring reactor inventory after scram is ensured automatically based on the reactor output before scram of a BWR type reactor. That is, if scram should occur, a feedwater flow rate just before the scram is stored by reactor output signals. Further, the amount of feedwater required after the scram is determined based on the output of the memory. The reactor power after the scram based on a feedwater flow rate and a main steam flow rate is inputted to an integrator, to calculate and output the amount of the feedwater flow rate (1) injected after the scram for the inventory. A coast down flowrate (2) in a case of pump trip is forecast by the output signals. Automatic trip is outputted to all turbine driving feedwater pumps when the sum of (1) and (2) exceeds a necessary and sufficient amount of feedwater required for ensuring inventory. For motor driving feedwater pumps, only a portion, for example, one of the pumps is automatically started while other pumps are stopped their operation, only in this case, to prevent excess water feeding. (I.S.)

  13. Autonomous Control of Space Nuclear Reactors Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Nuclear reactors to support future lunar and Mars robotic and manned missions impose new and innovative technological requirements for their control and protection...

  14. Reactor water level control device and water level control method

    International Nuclear Information System (INIS)

    Kawamura, Shin-ichiro; Habuka, Minoru.

    1996-01-01

    The present invention provide a device of and a method for controlling fluctuation of reactor water level caused upon insertion of selected control rods performed in order to ensure reactor stability when recycling pumps of a BWR-type power plant are stopped. When it is detected that reactor operation determined by reactor power and reactor core flow rate reaches unstable region after stoppage of a portion of the recycling pumps, the speed of integral recycling pumps is lowered to a predetermined speed of revolution conducted together with the insertion of selected control rods to suppress fluctuation of the reactor water level. Namely, (1) a recycling pump state monitoring device receives recycling pump state signals as an input and outputs recycling pump stopping detection signals. (2) A selected control rod insertion operation demand judging device judges the insurance of reactor stability due to insertion of the selected control rods and outputs selected control rod insertion operation demand signals. (3) A recycling pump speed control device outputs recycling pump speed control signals to control reactor core flow rate. (4) A minimum flow rate control valve controlling device outputs minimum flow rate control valve opening operation demand signals after a predetermined period of time. (5) A feed water pump minimum flow rate control valve is disposed to a feed water pump bypass channel and operated by the output signals of the device (4). (I.S.)

  15. Complete automation of nuclear reactors control

    International Nuclear Information System (INIS)

    Weill, J.

    1955-01-01

    The use of nuclear reactor for energy production induces the installation of automatic control systems which need to be safe enough and can adapt to the industrial scale of energy production. These automatic control systems have to insure the constancy of power level and adjust the power produced to the energy demand. Two functioning modes are considered: nuclear plant connected up to other electric production systems as hydraulic or thermic plants or nuclear plants functioning on an independent network. For nuclear plants connected up with other production plants, xenon poisoning and operating cost lead to keep working at maximum power the nuclear reactors. Thus, the power modulation control system will not be considered and only start-up control, safety control, and control systems will be automated. For nuclear power plants working on an independent network, the power modulation control system is needed to economize fuel. It described the automated control system for reactors functioning with constant power: a power measurement system constituted of an ionization chamber and a direct-current amplifier will control the steadfastness of the power produced. For reactors functioning with variable power, the automated power control system will allow to change the power and maintain it steady with all the necessary safety and will control that working conditions under P max and R max (maximum power and maximum reactivity). The effects of temperature and xenon poisoning will also be discussed. Safety systems will be added to stop completely the functioning of the reactor if P max is reached. (M.P.)

  16. Development of digital plant protection system for Korean Next Generation Reactor

    International Nuclear Information System (INIS)

    Suk-Joon Park

    1998-01-01

    A Digital Plant Protection System (DPPS) for Korean Next Generation Reactor (KNGR) is being developed using the Programmable Logic Controller (PLC) technology. For the design verification, the development of the DPPS prototype is progressing at this time. The prototype hardware equipment is installed and software coding is started. DPPS software is being coded by strict software V and V activities and function block language that uses simple graphical symbols. By adopting the PLC technology, the design of DPPS is possible to take full advantages in areas such as automatic testing, simplified calibration, improved isolation between redundant channels, reduced internal and external wiring and increased plant availability. (author)

  17. Control system for a small fission reactor

    Science.gov (United States)

    Burelbach, J.P.; Kann, W.J.; Saiveau, J.G.

    1985-02-08

    A system for controlling the reactivity of a small fission reactor includes an elongated, flexible hollow tube in the general form of a helical coiled spring axially positioned around and outside of the reactor vessel in an annular space between the reactor vessel and a surrounding cylindrical-shaped neutron reflector. A neutron absorbing material is provided within the hollow tube with the rate of the reaction controlled by the extension and compression of the hollow tube, e.g., extension of the tube increases reactivity while its compression reduces reactivity, in varying the amount of neutron absorbing material disposed between the reactor vessel and the neutron reflector. Conventional mechanical displacement means may be employed to control the coil density of the hollow tube as desired.

  18. Control aid for xenon vibration in reactor

    International Nuclear Information System (INIS)

    Kanekawa, Takashi.

    1990-01-01

    In the present invention, the control operation for suppressing xenon vibrations in a reactor is aided for saving forecasting analysis and operator's skills. That is, parameters to be controlled for the suppression of xenon vibrations are power distribution, iodine distribution and xenon distribution. But what can be observed by operaters by the conventional fast overtone method is only the output distribution. In the present invention, the output level of the reactor core is always observed. Then, mathematical processings are conducted for the iodine distribution, the xenon distribution and the power distribution in the reactor core based on the histeresis of the parameters obtained by the measurement using physical constants and reactor design data. The xenon vibration control is aided by displaying the change with time of the distortion in axial direction. Accordingly, operators can always recognize the axial distortion of the power distribution, the iodine distribution and the xenon distribution. (I.S.)

  19. Nuclear reactor shutdown control rod assembly

    International Nuclear Information System (INIS)

    Bilibin, K.

    1988-01-01

    This patent describes a nuclear reactor having a reactor core and a reactor coolant flowing therethrough, a temperature responsive, self-actuated nuclear reactor shutdown control rod assembly, comprising: an upper drive line terminating at its lower end with a substantially cylindrical wall member having inner and outer surfaces; a lower drive line having a lower end adapted to be attached to a neutron absorber; a ring movable disposed about the outer surface of the wall member of the upper drive line; thermal actuation means adapted to be in heat exchange relationship with coolant in an associated reactor core and in contact with the ring, and balls located within the openings in the upper drive line. When reactor coolant approaches a predetermined design temperature the actuation means moves the ring sufficiently so that the balls move radially out from the recess and into the space formed by the second portion of the ring thereby removing the vertical support for the lower drive line such that the lower drive line moves downwardly and inserts an associated neutron absorber into an associated reactor core resulting in automatic reduction of reactor power

  20. Digital control for turbogas units; Control digital para unidades turbogas

    Energy Technology Data Exchange (ETDEWEB)

    Garcia Beltran, Carlos Daniel

    1997-02-01

    The present thesis deals with the rehabilitation of the control system for the gas turbines W501 of the Gomez Palacio Combined Cycle Power Station in the state of Durango, Mexico. The first part of the development deals with a re-engineering process of software applied to the digital control system of the gas turbines of the Gomez Palacio Combined Cycle Power Station. This process was developed using concepts of several branches of engineering: a) involved the knowledge of the software engineering, using formal methods for the analysis of the original system and the redesign of the new system; b) The control engineering was used in the analysis of diverse control and automation strategies employed for gas turbines control, with the objective of verifying the type of instructions and existing routines within the software. The final product of this stage is a modulated programmatic system, based on structured design that is functionally a mirror image of the original system. The system obtained conformed by five main modules which are based on a model proposed originally for control by batch: i) Man Machine Interface, ii) Regulatory Control, iii) Protections, iv) Logic sequences and v) Supervision. The second stage of development was the improvement of the speed control of the turbine. When a turbogas unit is controlled, it must be taken into account several operation stages such as the starting, the control in stable state and the shut down. The real behavior of the turbine during the starting, and mainly the great number of backward movements produced, proposed by itself the search of a new controller who more closely maintained the acceleration specifications whereupon the turbine was designed. The development of a new control algorithm began with the analysis of the process, trying to identify which are the critical stages of this one and be able to evaluate in an objective form the advantages of an algorithm upon the other. It was continued with the

  1. Advanced Reactor Licensing: Experience with Digital I&C Technology in Evolutionary Plants

    Energy Technology Data Exchange (ETDEWEB)

    Wood, RT

    2004-09-27

    This report presents the findings from a study of experience with digital instrumentation and controls (I&C) technology in evolutionary nuclear power plants. In particular, this study evaluated regulatory approaches employed by the international nuclear power community for licensing advanced l&C systems and identified lessons learned. The report (1) gives an overview of the modern l&C technologies employed at numerous evolutionary nuclear power plants, (2) identifies performance experience derived from those applications, (3) discusses regulatory processes employed and issues that have arisen, (4) captures lessons learned from performance and regulatory experience, (5) suggests anticipated issues that may arise from international near-term deployment of reactor concepts, and (6) offers conclusions and recommendations for potential activities to support advanced reactor licensing in the United States.

  2. Control Rod Malfunction at the NRAD Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Thomas L. Maddock

    2010-05-01

    The neutron Radiography Reactor (NRAD) is a training, research, and isotope (TRIGA) reactor located at the INL. The reactor is normally shut down by the insertion of three control rods that drop into the core when power is removed from electromagnets. During a routine shutdown, indicator lights on the console showed that one of the control rods was not inserted. It was initially thought that the indicator lights were in error because of a limit switch that was out of adjustment. Through further testing, it was determined that the control rod did not drop when the scram switch was initially pressed. The control rod anomaly led to a six month shutdown of the reactor and an in depth investigation of the reactor protective system. The investigation looked into: scram switch operation, console modifications, and control rod drive mechanisms. A number of latent issues were discovered and corrected during the investigation. The cause of the control rod malfunction was found to be a buildup of corrosion in the control rod drive mechanism. The investigation resulted in modifications to equipment, changes to both operation and maintenance procedures, and additional training. No reoccurrences of the problem have been observed since corrective actions were implemented.

  3. Centralized digital control of accelerators

    International Nuclear Information System (INIS)

    Melen, R.E.

    1984-01-01

    Upon careful examination of the architecture of SLAC's computer control systems, it becomes evident that the distribution of the systems' intelligence generally falls into tree-like layers. The first layer typically consists of a central computer complex incorporating one or more relatively large and powerful processors. The more modern systems use state-of-the-art 32-bit processors with several megabytes of RAM and several hundreds of megabytes of disk memory. Further, they support extensive user-friendly operating systems and program development facilities. The second layer typically consists of several smaller processors which are downloaded from the central complex and whose primary task is to provide data acquisition and distribution. The more modern systems are 16-bit processors with several hundred kilobytes of RAM and no disk memory. The third layer typically consists of several tens or hundreds of micro-processors, each dedicated to a single device. The micro-processors for these ''dedicated intelligent controllers'' are small and inexpensive and typically require less than 32 kilobytes of RAM or EPROM memory. Their hardware may be general purpose in nature or may be built into the architecture of the device itself. Figure 5 illustrates several of the relevant features of each of these layers. This paper serves to illustrate that SLAC is commited to the centralized digital control of its accelerators

  4. Heating control system for nuclear reactor

    International Nuclear Information System (INIS)

    Shinohara, Kaoru.

    1981-01-01

    Purpose: To automatically control reactor heating while keeping the condition of temperature rising rate by determining the deviations based on the reactor water temperature, the aimed temperature and the aimed temperature rising rate and operating control rods. Constitution: Actual temperature in the reactor is measured by a temperature detector and compared with a value from a setter to determine the temperature deviation. While on the other hand, the rising rate for the measured temperature is calculated in a differentiator and compared with a value from a setter to determine the deviation, which is passed through an integrator to calculate the deviation for the temperature rising rate. The signals for the temperature deviation and the temperature rising rate deviation are selected in a lower value preference circuit and the operation amount for the control rod is judged in a control rod operation judging section depending on the deviation amount. The control rod to be operated is determined in a sequence control section for the selection of control rod. The control rod selected and the direction of the operation are displayed on a display and the selected control rod is automatically driven by a control rod drives to thereby carry our reactor heating. (Furukawa, Y.)

  5. Rolls-Royce digital Rod Control System

    International Nuclear Information System (INIS)

    Pouillot, M.

    2010-01-01

    Full text of publication follows: Rolls-Royce has developed a new generation of Rod Control System, based on 40 years of experience. The fifth-generation Rod Control System (RCS) from Rolls-Royce offers a reliable, modular design with adaptability to your preferred platform, for modernization projects or new reactors. Flexible implementation provides the option for you to keep existing cabinets, which permits you to optimize installation approach. Main features for the power part: - Control Rod Drive Mechanism (CRDM) type: 3-coil. - Independent control of each sub-bank. - Each sub-bank is controlled by a cycler unit and 3 identical power racks, each including 4 identical power modules and a common power-supply module. - Coil-per-coil digital control: each power module embeds power-conversion, current-control, and current-monitoring functions for one coil. Control and monitoring are carried out by separate electronics in the module. Current is digitized and fully monitored by means of min-max templates. - A double-hold function is included: a power module assigned to a gripper will activate its coil if a fault risking to cause a reactor trip occurs. - Power modules are standardized, hot-pluggable and self-configured: a power module includes a set of parameters for each type of coil SG, MG, LC. The module recognizes the rack it is plugged in, and chooses automatically parameters to be used. Main benefits: - Reduced operational, maintenance, training, and inventory costs: standardization of power modules and integration of control and monitoring on the same PC-card lead to a drastic reduction of spare part types, and simplification of the system. - Easy maintenance: - Replacement of a power module solves nearly all failures due to current control or monitoring for a coil. It is done instantly thanks to hot-plug capability. - On the front plate of power-modules, LEDs provide useful information for diagnostic: current setpoint from cycler, output current bar

  6. Multivariable Feedback Control of Nuclear Reactors

    Directory of Open Access Journals (Sweden)

    Rune Moen

    1982-07-01

    Full Text Available Multivariable feedback control has been adapted for optimal control of the spatial power distribution in nuclear reactor cores. Two design techniques, based on the theory of automatic control, were developed: the State Variable Feedback (SVF is an application of the linear optimal control theory, and the Multivariable Frequency Response (MFR is based on a generalization of the traditional frequency response approach to control system design.

  7. Nuclear reactor plants and control systems therefor

    International Nuclear Information System (INIS)

    de Boer, G.A.; de Hex, M.

    1976-01-01

    A nuclear reactor plant is described comprising at least two hydraulically separated but thermally interconnected heat conveying circuits, of which one is the reactor circuit filled with a non-water medium and the other one is the water-steam-circuit equipped with a steam generator, a feed water conduit controlled by a valve and a steam turbine, and a control system mainly influenced by the pressure drop caused in said feed water conduit and its control valve and having a value of at least 10 bars at full load

  8. Light Water Reactor Sustainability Program: Digital Architecture Project Plan

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Ken [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-09-01

    There are many technologies available to the nuclear power industry to improve efficiency in plant work activities. These range from new control room technologies to those for mobile field workers. They can make a positive impact on a wide range of performance objectives – increase in productivity, human error reduction, validation of results, accurate transfer of data, and elimination of repetitive tasks. It is expected that the industry will more and more turn to these technologies to achieve these operational efficiencies to lower costs. At the same time, this will help utilities manage a looming staffing problem as the inevitable retirement wave of the more seasoned workers affects both staffing levels and knowledge retention. A barrier to this wide-scale implementation of new technologies for operational efficiency is the lack of a comprehensive digital architecture that can support the real-time information exchanges needed to achieve the desired operational efficiencies. This project will define an advanced digital architecture that will accommodate the entire range of system, process, and plant worker activity to enable the highest degree of integration, thereby creating maximum efficiency and productivity. This pilot project will consider a range of open standards that are suitable for the various data and communication requirements of a seamless digital environment. It will map these standards into an overall architecture to support the II&C developments of this research program.

  9. Conversion of a servomanipulator from analog to digital control

    International Nuclear Information System (INIS)

    Killough, S.M.; Martin, H.L.; Hamel, W.R.

    1986-01-01

    Oak Ridge National Laboratory (ORNL) has developed expertise in computer control of force-reflecting master/slave servomanipulators as a result of research for the Consolidated Fuel Reprocessing Program. These computer control capabilities have been applied to a commercially available servomanipulator, the TeleOperator Systems SM-229. All of the servo drive and control circuitry has been replaced with commercially available digital controls and amplifiers, and a customer software - package has been developed at ORNL. This conversion to digital computer control resulted in significant improvements in force-reflection characteristics, ease of operation, diagnostic capabilities, indexing features, and potential increased reliability. The system will be used at the Tokamak Fusion Test Reactor at the Princeton Plasma Physics Laboratory (PPPL) for maintenance demonstrations

  10. Power control device in nuclear reactor

    International Nuclear Information System (INIS)

    Koyama, Kazuaki.

    1981-01-01

    Purpose: To enable smooth power changes in power conditioning systems by calculating forecast values for the neutron flux distribution and power distribution and by controlling the driving speed of control rods so as to correspond the forecast values with aimed values. Constitution: Control rod position is detected by a position detector and sent to a control computer as the position information. At the same time, the neutron flux distribution information is obtained by the neutron monitors, the power distribution information is obtained by a reactor power computer and they are outputted to the control computer. The control computer calculates the forecast values for the neutron flux distribution and the reactor power distribution from the information, and compares them with the aimed values from a setter and then outputs control signals so as to correspond the forecast values with the aimed values. The control rods can be inserted in appropriate velocity by the control signals. (Horiuchi, T.)

  11. Autonomous Control of Space Reactor Systems

    International Nuclear Information System (INIS)

    Belle R. Upadhyaya; K. Zhao; S.R.P. Perillo; Xiaojia Xu; M.G. Na

    2007-01-01

    Autonomous and semi-autonomous control is a key element of space reactor design in order to meet the mission requirements of safety, reliability, survivability, and life expectancy. Interrestrial nuclear power plants, human operators are available to perform intelligent control functions that are necessary for both normal and abnormal operational conditions

  12. Autonomous Control of Space Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Belle R. Upadhyaya; K. Zhao; S.R.P. Perillo; Xiaojia Xu; M.G. Na

    2007-11-30

    Autonomous and semi-autonomous control is a key element of space reactor design in order to meet the mission requirements of safety, reliability, survivability, and life expectancy. Interrestrial nuclear power plants, human operators are avilable to perform intelligent control functions that are necessary for both normal and abnormal operational conditions.

  13. Reliability Estimation for Digital Instrument/Control System

    International Nuclear Information System (INIS)

    Yang, Yaguang; Sydnor, Russell

    2011-01-01

    Digital instrumentation and controls (DI and C) systems are widely adopted in various industries because of their flexibility and ability to implement various functions that can be used to automatically monitor, analyze, and control complicated systems. It is anticipated that the DI and C will replace the traditional analog instrumentation and controls (AI and C) systems in all future nuclear reactor designs. There is an increasing interest for reliability and risk analyses for safety critical DI and C systems in regulatory organizations, such as The United States Nuclear Regulatory Commission. Developing reliability models and reliability estimation methods for digital reactor control and protection systems will involve every part of the DI and C system, such as sensors, signal conditioning and processing components, transmission lines and digital communication systems, D/A and A/D converters, computer system, signal processing software, control and protection software, power supply system, and actuators. Some of these components are hardware, such as sensors and actuators, their failure mechanisms are well understood, and the traditional reliability model and estimation methods can be directly applied. But many of these components are firmware which has software embedded in the hardware, and software needs special consideration because its failure mechanism is unique, and the reliability estimation method for a software system will be different from the ones used for hardware systems. In this paper, we will propose a reliability estimation method for the entire DI and C system reliability using a recently developed software reliability estimation method and a traditional hardware reliability estimation method

  14. Control rod for FBR type reactor

    International Nuclear Information System (INIS)

    Nakai, Koichi.

    1993-01-01

    In a control rod for an LMFBR type reactor, a thermal resistor is disposed between a temperature sensitive cylinder and a cam unit support rod. A thermal expansion difference due to the temperature difference is caused between the temperature sensitive cylinder and the cam unit support rod only upon abrupt temperature change of coolants. A control rod shaft extending mechanism of downwardly depressing an absorbent portion by amplifying the thermal expansion difference by an extension link mechanism and the cam unit is provided. The thermal resistor comprises inconel 625 or like other steel of small heat conductivity. If a certain abnormality should cause to the reactor system to elevate the coolant temperature in the reactor elevates abruptly and the reactor shutdown system does not actuate, since the control rod extension shaft extends to urge the absorbent and lower the reactor core reactivity, so that leading to serious accident can be prevented surely. Further, the control rod extension shaft does not extend upon moderate temperature elevation in the usual startup and causes no unnecessary reactivity change. (N.H.)

  15. Application of digital process controller for automatic pulse operation in the NSRR

    International Nuclear Information System (INIS)

    Ishijima, K.; Ueda, T.; Saigo, M.

    1992-01-01

    The NSRR at JAERI is a modified TRIGA Reactor. It was built for investigating reactor fuel behavior under reactivity initiated accident (RIA) conditions. Recently, there has been a need to improve the flexibility of pulsing operations in the NSRR to cover a wide range of accidental situations, including RIA events at elevated power levels, and various abnormal power transients. To satisfy this need, we developed a new reactor control system which allows us to perform 'Shaped Pulse Operation: SP' and 'Combined Pulse Operation: CP'. Quick, accurate and complicated manipulation of control rods was required to realize these operations. Therefore we installed a new reactor control system, which we call an automatic pulse control system. This control system is composed of digital processing controllers and other digital equipments, and is fully automated and highly accurate. (author)

  16. Nuclear reactor kinetics and plant control

    CERN Document Server

    Oka, Yoshiaki

    2013-01-01

    Understanding time-dependent behaviors of nuclear reactors and the methods of their control is essential to the operation and safety of nuclear power plants. This book provides graduate students, researchers, and engineers in nuclear engineering comprehensive information on both the fundamental theory of nuclear reactor kinetics and control and the state-of-the-art practice in actual plants, as well as the idea of how to bridge the two. The first part focuses on understanding fundamental nuclear kinetics. It introduces delayed neutrons, fission chain reactions, point kinetics theory, reactivit

  17. Digital Systems Implemented at the IPEN Nuclear Research Reactor (IEA-R1): Results and Necessities

    International Nuclear Information System (INIS)

    Nahuel-Cardenas, Jose-Patricio; Madi-Filho, Tufic; Ricci-Filho, Walter; Rodrigues-de-Carvalho, Marcos; Lima-Benevenuti, Erion-de; Gomes-Neto, Jose

    2013-06-01

    obsolescence of some electrical and electronic systems. In this work we will show a retrospective and results of digital systems applied to IEA-R1 reactor concerning electronic equipments and systems refurbishment and modernization and the necessity of a new control console implementation. (authors)

  18. A digital control system for neutron spectrometers

    DEFF Research Database (Denmark)

    Hansen, Knud Bent; Skaarup, Per

    1968-01-01

    A description is given of the principles of a digital system used to control neutron spectrometers. The system is composed of independent functional units with the control programme stored on punched paper tape or in a computer.......A description is given of the principles of a digital system used to control neutron spectrometers. The system is composed of independent functional units with the control programme stored on punched paper tape or in a computer....

  19. The Calculation Process of Channel Uncertainty on Digital System for Integral Reactor

    International Nuclear Information System (INIS)

    Moon, Hee Gun; Kim, Sung Hun; Kim, Jung Seon; Park, Heui Youn; Koo, In Soo

    2005-01-01

    The Channel Uncertainty is very important factor on the Safety Analysis input data because the channel uncertainty is the input data for calculating the trip setpoint and verify the hardware accuracy. But, the instrumentation system is changed to digital system due to technical improvement in these days. So, there are occurred to problems when digital system adopt directly the calculation method of uncertainty on analog system. This paper shows that the difference of the calculation process between analog system and digital system. And also, it presents consideration for calculating the uncertainty on digital system and the calculation process of uncertainty on digital system for Integral Reactor

  20. Control system studies for thermionic reactors

    Science.gov (United States)

    Hermsen, R. J.; Gronroos, H. G.

    1978-01-01

    In core thermionic reactor concepts are of interest for space missions that require electrical power in the range of a few tens of kilowatts up to several megawatts. The physical principle involved--thermionic direct conversion of heat to electricity at net efficiencies up to 15 percent--offers potential advantages when compared to other nuclear powerplant concepts. However, the integration of the thermionic diode electrode structure with high-temperature nuclear fuel materials presents new design problems and new reactor physical constraints. Among the topics that must be investigated are those associated with the control system. The results of analytical and simulation studies of thermionic reactor control performed at the Jet Propulsion Laboratory are discussed.

  1. An autonomous control framework for advanced reactors

    Directory of Open Access Journals (Sweden)

    Richard T. Wood

    2017-08-01

    Full Text Available Several Generation IV nuclear reactor concepts have goals for optimizing investment recovery through phased introduction of multiple units on a common site with shared facilities and/or reconfigurable energy conversion systems. Additionally, small modular reactors are suitable for remote deployment to support highly localized microgrids in isolated, underdeveloped regions. The long-term economic viability of these advanced reactor plants depends on significant reductions in plant operations and maintenance costs. To accomplish these goals, intelligent control and diagnostic capabilities are needed to provide nearly autonomous operations with anticipatory maintenance. A nearly autonomous control system should enable automatic operation of a nuclear power plant while adapting to equipment faults and other upsets. It needs to have many intelligent capabilities, such as diagnosis, simulation, analysis, planning, reconfigurability, self-validation, and decision. These capabilities have been the subject of research for many years, but an autonomous control system for nuclear power generation remains as-yet an unrealized goal. This article describes a functional framework for intelligent, autonomous control that can facilitate the integration of control, diagnostic, and decision-making capabilities to satisfy the operational and performance goals of power plants based on multimodular advanced reactors.

  2. An autonomous control framework for advanced reactors

    Energy Technology Data Exchange (ETDEWEB)

    Wood, Richard T.; Upadhyaya, Belle R.; Floyd, Dan C. [Dept. of Nuclear Engineering, University of Tennessee, Knoxville (United States)

    2017-08-15

    Several Generation IV nuclear reactor concepts have goals for optimizing investment recovery through phased introduction of multiple units on a common site with shared facilities and/or reconfigurable energy conversion systems. Additionally, small modular reactors are suitable for remote deployment to support highly localized microgrids in isolated, underdeveloped regions. The long-term economic viability of these advanced reactor plants depends on significant reductions in plant operations and maintenance costs. To accomplish these goals, intelligent control and diagnostic capabilities are needed to provide nearly autonomous operations with anticipatory maintenance. A nearly autonomous control system should enable automatic operation of a nuclear power plant while adapting to equipment faults and other upsets. It needs to have many intelligent capabilities, such as diagnosis, simulation, analysis, planning, reconfigurability, self-validation, and decision. These capabilities have been the subject of research for many years, but an autonomous control system for nuclear power generation remains as-yet an unrealized goal. This article describes a functional framework for intelligent, autonomous control that can facilitate the integration of control, diagnostic, and decision-making capabilities to satisfy the operational and performance goals of power plants based on multimodular advanced reactors.

  3. Advanced reactor instrumentation and control reliability and risk assessment

    International Nuclear Information System (INIS)

    Fullwood, R.; Gunther, W.; Valente, J.; Azarm, M.A.

    1991-01-01

    Advanced nuclear power reactors will used different approaches to achieving a higher level of safety than the first generation. One approach used the technological developments in computation and electronics in the form of digital instrumentation and control (I ampersand C) to enhance the reliability, and accuracy of information for plant control, responding to the information, and controlling the plant and its systems under normal and upset environments in various states of degradation. Evaluating the reliability and safety of advanced I ampersand C systems requires determining the reliability of the I ampersand C used in the advanced reactors which involves distributed processing, data pile-up, interactive systems, the man-machine interface, various forms of automatic control, and systems interactions. From these analyses will come an understanding of the potential of the new I ampersand C, and protection from its vulnerabilities to enhance the safe operation of the new plants. Technological, safety, reliability, and regulatory issues associated with advanced I ampersand C for the new reactors are discussed herein. The issues are presented followed by suggested approaches to their resolution

  4. Magnetic switch for reactor control rod. [LMFBR

    Science.gov (United States)

    Germer, J.H.

    1982-09-30

    A magnetic reed switch assembly is described for activating an electromagnetic grapple utilized to hold a control rod in position above a reactor core. In normal operation the magnetic field of a permanent magnet is short-circuited by a magnetic shunt, diverting the magnetic field away from the reed switch. The magnetic shunt is made of a material having a Curie-point at the desired release temperature. Above that temperature the material loses its ferromagnetic properties, and the magnetic path is diverted to the reed switch which closes and short-circuits the control circuit for the control rod electro-magnetic grapple which allows the control rod to drop into the reactor core for controlling the reactivity of the core.

  5. Advanced Digital Controller Development Program, Task I.

    Science.gov (United States)

    1982-07-01

    digital controller through an electra-hydraulic servo valve ( EHSV ) and a position transducer. A pair of hydraulic cylinders will actuate a bell crank...AD-A129 269 ADVANCED DIGITAL CONTROLLER DEVELOPMENT PROGRAM TASK 1 1/1 ()GENERAL MOTORS CORP INDIANAPOLIS IN DETROIT DIESEL ALLISON D0. d H HUNTER JL...BUREAU OF S IANDARDS I-63 A I-16 2 I "-I [ I1 I ? I I I I II I I AFWAL-TR-82-2047 ’ Advanced Digital Controller < Development Program, Task I Detroit

  6. Minimization of PWR reactor control rods wear

    International Nuclear Information System (INIS)

    Ponzoni Filho, Pedro; Moura Angelkorte, Gunther de

    1995-01-01

    The Rod Cluster Control Assemblies (RCCA's) of Pressurized Water Reactors (PWR's) have experienced a continuously wall cladding wear when Reactor Coolant Pumps (RCP's) are running. Fretting wear is a result of vibrational contact between RCCA rodlets and the guide cards which provide lateral support for the rodlets when RCCA's are withdrawn from the core. A procedure is developed to minimize the rodlets wear, by the shuffling and axial reposition of RCCA's every operating cycle. These shuffling and repositions are based on measurement of the rodlet cladding thickness of all RCCA's. (author). 3 refs, 2 figs, 2 tabs

  7. Nuclear reactor control room construction

    Science.gov (United States)

    Lamuro, R.C.; Orr, R.

    1993-11-16

    A control room for a nuclear plant is disclosed. In the control room, objects labelled 12, 20, 22, 26, 30 in the drawing are no less than four inches from walls labelled 10.2. A ceiling contains cooling fins that extend downwards toward the floor from metal plates. A concrete slab is poured over the plates. Studs are welded to the plates and are encased in the concrete. 6 figures.

  8. Nuclear reactor control room construction

    Science.gov (United States)

    Lamuro, Robert C.; Orr, Richard

    1993-01-01

    A control room 10 for a nuclear plant is disclosed. In the control room, objects 12, 20, 22, 26, 30 are no less than four inches from walls 10.2. A ceiling 32 contains cooling fins 35 that extend downwards toward the floor from metal plates 34. A concrete slab 33 is poured over the plates. Studs 36 are welded to the plates and are encased in the concrete.

  9. Nuclear reactor control room construction

    International Nuclear Information System (INIS)

    Lamuro, R.C.; Orr, R.

    1993-01-01

    A control room for a nuclear plant is disclosed. In the control room, objects labelled 12, 20, 22, 26, 30 in the drawing are no less than four inches from walls labelled 10.2. A ceiling contains cooling fins that extend downwards toward the floor from metal plates. A concrete slab is poured over the plates. Studs are welded to the plates and are encased in the concrete. 6 figures

  10. Humidity control device in a reactor container

    International Nuclear Information System (INIS)

    Aizawa, Motohiro; Igarashi, Hiroo; Osumi, Katsumi; Kimura, Takashi.

    1986-01-01

    Purpose: To provide a device capable of maintaining the inside of a container under high humidity circumstantial conditions causing less atmospheric corrosions, in order to prevent injuries due to atmospheric corrosions to smaller diameter stainless steel pipeways in the reactor container. Constitution: Stress corrosion cracks (SCC) to the smaller diameter stainless steel pipeways are caused dependent on the relative humidity and it is effective as the countermeasure against SCC to maintain the relative humidity at a low level less than 30 % or high level greater than 60 %. Based on the above findings, a humidity control device is disposed so as to maintain the relative humidity for the atmosphere within a reactor core on a higher humidity region. The device is adapted such that recycling gas in the dry-well coolant circuit is passed through an orifice to atomize the water introduced from feedwater pipe and introduce into a reactor core or such that the recycling gases in the dry-well cooling circuit are bubbled into water to remove chlorine gas in the reactor container gas thereby increasing the humidity in the reactor container. (Kamimura, M.)

  11. Damper mechanism for nuclear reactor control elements

    International Nuclear Information System (INIS)

    Taft, W.E.

    1976-01-01

    A damper mechanism which provides a nuclear reactor control element decelerating function at the end of the scram stroke is described. The total damping function is produced by the combination of two assemblies, which operate in sequence. First, a tapered dashram assembly decelerates the control element to a lower velocity, after which a spring hydraulic damper assembly takes over to complete the final damping. 3 claims, 2 figures

  12. Model predictive controller design of hydrocracker reactors

    OpenAIRE

    GÖKÇE, Dila

    2014-01-01

    This study summarizes the design of a Model Predictive Controller (MPC) in Tüpraş, İzmit Refinery Hydrocracker Unit Reactors. Hydrocracking process, in which heavy vacuum gasoil is converted into lighter and valuable products at high temperature and pressure is described briefly. Controller design description, identification and modeling studies are examined and the model variables are presented. WABT (Weighted Average Bed Temperature) equalization and conversion increase are simulate...

  13. Method of controlling reactor powers

    International Nuclear Information System (INIS)

    Takayama, Yoshito.

    1980-01-01

    Purpose: To compensate power fluctuations with no distortions in the power distribution by varying the temperature of heavy water as moderators to thereby vary the density of boron therein. Method: A value set for the expected power is compared with a detected value for the power and the power deviation is inputted into a judging circuit. If the deviation exceeds a predetermined value, a power fluctuation amount signal from the judging circuit is inputted into a function generator, where the heavy water temperature fluctuation amount corresponding to the reactivity fluctuation amount is computed and outputted. The heavy water temperature fluctuation amount signal is compared with the detected heavy water temperature and a temperature deviation signal is inputted into a function generator, where the opening degree for a control valve is computed and an opening degree signal is supplied to a temperature control valve. The control valve, upon receiving the signal, regulates the amount of coolants to control the temperature of heavy water. (Sekiya, K.)

  14. Comparative assessment of instrumentation and control (I and C) system architectures for research reactors

    International Nuclear Information System (INIS)

    Khalil, Rah Man; Heo, Gyun Young; Son, Han Seong; Kim, Young Ki; Park, Jae Kwan

    2012-01-01

    Application of digital I and C has increased in nuclear industry since last two decades but lack of experience, innovative and naive nature of technology and insufficient failure information raised questions on its use. The issues has been highlighted due to the use of digital I and C which were not relevant to analog. These are the potential weakness of digital systems for Common Cause Failure, threat to system security and reliability due to inter channel communication, need for highly integrated control room and difficulty to assess the digital I and C reliability. In the existing scenario, HANARO and JRTR have hybrid I and C systems (digital plus analog) whereas OPAL is fully digitalized. In order to authenticate the choice of fully digital I and C architecture for research reactor, it is required to perform assessment from risk point of view, cyber security as well other issues. The architecture assessment method and restrictions are discussed in the next part of article

  15. Operation control equipment for BWR type reactor

    International Nuclear Information System (INIS)

    Izumi, Masayuki; Takeda, Renzo.

    1981-01-01

    Purpose: To improve the temperature balance in a feedwater heater by obtaining the objective value of a feedwater enthalpy upon calculation of respective measured values and controlling the opening or closing of an extraction valve so that the objective value may coincide with the measured value, thereby averaging the axial power distribution. Constitution: A plurality of stages of extraction lines are connected to a turbine, and extraction valves are respectively provided at the lines. By calculating the measured values of ractor pressure, reactor core flow rate, vapor flow rate and reactor core inlet enthalpy determined to predetermined value using heat balance the objective feedwater enthalpy is obtained, is fed as an extraction valve opening or closing signal from a control equipment, the extraction stages of the turbine extraction are altered in accordance with this signal, and the feedwater enthalpy is controlled. (Sekiya, K.)

  16. Control rod for a reactor

    International Nuclear Information System (INIS)

    Natori, Hisahide.

    1975-01-01

    Object: To change arrangement and density of each layer of neutron absorber in the control rod and to render rotation by each layer possible, whereby the neutron absorber may be rotated to readily flatten power distribution. Structure: Neutron absorbers such as boron and carbide are filled into stainless steel pipes, which are peripherally arranged in a multi-layer fashion. Arrangement and density of the neutron absorber by each layer are changed and rotation by each layer is made possible, whereby surface area of the absorber or the like is changed to flatten power distribution. (Furukawa, Y.)

  17. Controlling a nuclear reactor with dropped control rods

    International Nuclear Information System (INIS)

    Mc Atee, K.R.; Alsop, B.H.

    1987-01-01

    A control system is described for a nuclear power plant including a reactor with a core having an upper portion and a lower portion and control rods which are inserted into and withdrawn from the core of the reactor vertically to control reactivity in the core. The system comprises: means to measure neutron flux separately in the upper portion and the lower portion of the reactor and to generate from such measurements a signal representative of axial distribution of power between the upper and lower portions of the reactor core; means to detect a dropped control rod in the reactor and to generate a dropped rod signal in response thereto; means to generate an axial power distribution limit signal representative of a critical axial power distribution for a dropped rod condition; means to compare the axial power distribution signal to the axial power distribution limit signal and to generate an axial power distribution out of limits signal when the axial power distribution signal exceeds the axial power distribution limit signal; and means responsive only to the presence of both the dropped rod signal and the axial power distribution out of limits signal to generate a signal for shutting the reactor down

  18. Power flow control using distributed saturable reactors

    Science.gov (United States)

    Dimitrovski, Aleksandar D.

    2016-02-13

    A magnetic amplifier includes a saturable core having a plurality of legs. Control windings wound around separate legs are spaced apart from each other and connected in series in an anti-symmetric relation. The control windings are configured in such a way that a biasing magnetic flux arising from a control current flowing through one of the plurality of control windings is substantially equal to the biasing magnetic flux flowing into a second of the plurality of control windings. The flow of the control current through each of the plurality of control windings changes the reactance of the saturable core reactor by driving those portions of the saturable core that convey the biasing magnetic flux in the saturable core into saturation. The phasing of the control winding limits a voltage induced in the plurality of control windings caused by a magnetic flux passing around a portion of the saturable core.

  19. Computer Controlled Chemical Micro-Reactor

    International Nuclear Information System (INIS)

    Mechtilde, Schaefer; Eduard, Stach; Adreas, Foitzik

    2006-01-01

    Chemical reactions or chemical equilibria can be influenced and controlled by several parameters. The ratio of two liquid ingredients, the so called reactants or educts, plays an important role in determining the end product and its yield. The reactants must be weighed and accordingly mixed with the conventional batch mode. If the reaction is done in a microreactor or in several parallel working micro-reactors, units for allotting the educts in appropriate quantities are required. In this report we present a novel micro-reactor that allows the constant monitoring of the chemical reaction via Raman spectroscopy. Such monitoring enables an appropriate feedback on the steering parameters for the PC controlled micro-pumps for the appropriate educt flow rate of both liquids to get optimised ratios of ingredients at an optimised total flow rate. The micro-reactors are the core pieces of the design and are easily removable and can therefore be changed at any time to adapt the requirements of the chemical reaction. One type of reactor consists of a stainless steel base containing small scale milled channels covered with anodically bonded Pyrex glass. Another type of reactor has a base of anisotropically etched silicon, and is also covered with anodically bonded Pyrex glass. The glass window allows visual observation of the initial phase interface of the two educts in the reaction channels by optical microscopy and does not affect, in contrast to infrared spectroscopy, the Raman spectroscopic signal for detection of the reaction kinetics. On the basis of a test reaction, we present non-invasive and spatially highly resolved in-situ reaction analysis using Raman spectroscopy measured along the reaction channel at different locations

  20. Reactor instrumentation and control design and performance simulation for SP-100

    Science.gov (United States)

    Meyer, R. A.; Alley, A. D.; Halfen, F. J.; Brynsvold, G. V.

    1987-01-01

    The SP-100 flight system will be launched with all primary and secondary lithium in the solid state. Once in orbit, the reactor will be brought critical and maintained at a low power level while the lithium is thawed out. Once the system is thawed out, the reactor power will be controlled to provide the energy source required by the power conversion system to meet the payload electrical power requirements. The Reactor Instrumentation and Control subsystem which includes the reactor control drives, instrumentation and the digital controller provides for the control of the nuclear subsystem to perform these operating maneuvers as well as providing for automatic shutdown and restart under certain off-normal conditions. The design and performance of this system are described.

  1. RIMACS, Reactor Inspection Main Control System

    International Nuclear Information System (INIS)

    2008-01-01

    1 - Description of program or function: RIMACS prepares for automatic inspection files on each inspection item for the reactor. These automatic inspection files provide the data to move RIROB (Reactor Inspection Robot) with laser by interpreting the coordinates of LASPO (Laser Positioner) and the laser detecting device of RIROB in three dimensional space. In addition, when RIROB arrives at the inspecting location, the files provide all values of the manipulator's motions to acquire the ultrasonic data. RIMACS provides various modules in order to perform these complex functions, and the functions are programmed on graphic user interface for the convenience of the user. RIMACS provides various functions, such as insertion of reactor production data, selection of the reactor for inspection, the creation of automatic inspection file, the selection of the inspection item, inspection simulation, and automatic inspection procedures. It also provides all other functions, which are necessary for the inspection, such as operating program download and manual control of LASPO and RIROB, the inspection simulation and the inspection status display by means of the graphic screen, and SODAS (ultra-Sonic Data Acquisition System) drive verification. 2 - Methods: Moving path and operation procedures for inspection robot are generated automatically with Kinematics algorithm. 3 - Restrictions on the complexity of the problem: A graphics display with MS-Window capability is required

  2. Cobalt-60 control in Ontario Hydro reactors

    International Nuclear Information System (INIS)

    Lacy, C.S.

    1988-01-01

    This paper discusses the impact of specifying reduced Cobalt-59 in the primary heat transport circuit materials of construction on the radiation fields developed around the primary circuit. An eight-fold reduction in steam generator radiation fields due to Cobalt-60 has been observed for two identical sets of reactors, one with and one without Cobalt-59 control. The comparison is between eight reactors at the Pickering Nuclear Generating Station (PNGS). Units 5 to 8 (PNGS-B) are identical to Units 1 to 4 (PNGS-A) except that PNGS-B has reduced impurity Cobalt-59 in the alloys of construction and a reduced use of stellite. The effects of chemistry control are also discussed

  3. Computerized control system of reactor charging machine

    International Nuclear Information System (INIS)

    Liptak, P.; Krajci, L.; Valo, J.; Bacek, M.

    1992-01-01

    Features of a new system of charging machine control are described, based on a PC AT computer. The main advantages of the new system include reduction of manual manipulations the machine and effective control and recording of the manipulations steps the machine. The system allows to control, check and record all manipulation sequences on the machine above the reactor or in the area of the storage pool. Suitable graphics provides optimal comfort for the machine operator or physicist during manipulation with the system. (author) 7 figs., 1 ref

  4. Test on the reactor with the portable digital reactivity meter for physical experiment

    International Nuclear Information System (INIS)

    Huang Liyuan

    2010-01-01

    Test must be performed on the zero power reactor During the development of portable digital reactivity meter for physical experiment, in order to check its measurement function and accuracy. It describes the test facility, test core, test methods, test items and test results. The test results show that the instrument satisfy the requirements of technical specification, and satisfy the reactivity measurement in the physical experiments on reactors. (authors)

  5. Study and application of digital physical start-up system for nuclear reactor

    International Nuclear Information System (INIS)

    Qu Ronghong; Li Baoxiang; Xu Xiaolin

    2004-01-01

    The digital physical start-up system for nuclear reactor is introduced. The system was used successfully in physical start-up experiment of 10 MW high-temperature gas-cooled reactor. It is proved practically that the system not only runs reliably and calculates both rapidly and correctly and relieves the loads of operators, but also has the better characters of monitoring and showing the real-time results of experiments than the analog systems. (author)

  6. Control rod controlling device of nuclear reactor

    International Nuclear Information System (INIS)

    Arita, Setsuo; Okido, Fumiyasu.

    1997-01-01

    The present invention concerns a control rod drive mechanism for use in a BWR, which does not apply undesired effects on monitoring of neutron instrumentation systems. Control rods are operated using an induction electric motor equipped with an electromagnetic brake as a driving source. The induction electric motor and the electromagnetic brake are driven by ON/OFF control. Since a switching element for driving the induction electric motor and the electromagnetic brake can be kept ON or OFF during control rod operation, electromagnetic noises are not generated during the operation of the control rods. Accordingly, the neutron instrumentation systems do not undergoing effects of electromagnetic noises during operation of control rods, and the neutron instrumentation system can accurately be monitored. (N.H.)

  7. Experience with safety assessment of digital upgrading of IandC in VVER type reactors

    International Nuclear Information System (INIS)

    Wach, D.; Mulka, B.; Schnuerer, G.

    1997-01-01

    The digital upgrading of IandC systems important to safety in WWER type reactors requires a broad expertise in various knowledge fields. The approach of the Institute for safety Technology to the qualification and categorization of safety-critical software systems is highlighted. The role of the Institute in the qualification of the Teleperm XS and the type testing of its components is described. The aspects of the safety assessment of digital IandC systems in WWER type reactors is discussed in some detail. (A.K.)

  8. Training simulator for nuclear power plant reactor control model and method

    International Nuclear Information System (INIS)

    Czerbuejewski, F.R.

    1975-01-01

    A description is given of a method and system for the real-time dynamic simulation of a nuclear power plant for training purposes, wherein a control console has a plurality of manual and automatic remote control devices for operating simulated control rods and has indicating devices for monitoring the physical operation of a simulated reactor. Digital computer means are connected to the control console to calculate data values for operating the monitoring devices in accordance with the control devices. The simulation of the reactor control rod mechanism is disclosed whereby the digital computer means operates the rod position monitoring devices in a real-time that is a fraction of the computer time steps and simulates the quick response of a control rod remote control lever together with the delayed response upon a change of direction

  9. Design and implementation of the control system for the new console of TRIGA-3-Salazar Reactor

    International Nuclear Information System (INIS)

    Gonzalez M, J.L.

    1994-01-01

    TRIGA-3-Salazar Reactor was set in operation in 1968 and the aging of its components has cause the increasing in the maintenance. In the presence of this, it becomes necessary to replace the reactor console using new technologies, considering the incorporation of a personal computer. The aim of this work is the design and construction of the equipment interfaces as well as the digital computer program for the automation and control of the TRIGA-3-Salazar Reactor by means of a personal computer. (Author)

  10. Position detection device for nuclear reactor control device

    International Nuclear Information System (INIS)

    Ara, Katsuyuki.

    1989-01-01

    The present invention concerns a position detection device for control rods in a PWR type nuclear reactor, and it is an object thereof to improve the reliability by remarkably decreasing the number of coils required for the position detection of the control rods. That is, there are provided rod members by the number of K each having a length substantially equal with the moving range of the control rod and moving interlocked with the control rod. The rod members by the number of K are each divided into M sections along the axial direction. Each of the divided sections is provided with different magnetic permeability an n-levels so as to correspond to the n-step number (0, 1, ----- n-1) and k-digits, n-step numbers are allocated to the divided M sections. K-digit and n-step number are detected by detection means base on the output of coils disposed to the rod-members by the number of K. As a result, the moving position of the control rod is detected. According to the present invention, it may be possible to dispose coils only by the number of the rod member in one section, not as in the usual case where each one coil is positioned to each of the sections. Accordingly, the constitution of the device can be simplified to facilitate the maintenance. (K.M.)

  11. Implementation of a digital feedwater control system at Dresden Nuclear Power Plant, Units 2 and 3: Final report

    International Nuclear Information System (INIS)

    Zapotocky, A.; Popovic, J.R.; Fournier, R.D.

    1988-12-01

    This report describes the Digital Feedwater Control System Implementation at the Dresden 2 or 3 Units of the BWR Nuclear Power Plant owned by the Commonwealth Edison Company. The digital system has been operational in Unit 3 since August 1986, and in Unit 2 since April 1987. The Bailey Control's Network 90 based digital control system replaced the obsolete GE/MAC 5000 analog control system in the reactor feedwater control loop as a ''like-for-like'' replacement. Operational experience from the Digital Feedwater Control installations has been good and the system demonstrated better performance than the old analog systems. 14 refs., 15 figs., 17 tabs

  12. HIGH STRENGTH CONTROL RODS FOR NEUTRONIC REACTORS

    Science.gov (United States)

    Lustman, B.; Losco, E.F.; Cohen, I.

    1961-07-11

    Nuclear reactor control rods comprised of highly compressed and sintered finely divided metal alloy panticles and fine metal oxide panticles substantially uniformly distributed theretbrough are described. The metal alloy consists essentially of silver, indium, cadmium, tin, and aluminum, the amount of each being present in centain percentages by weight. The oxide particles are metal oxides of the metal alloy composition, the amount of oxygen being present in certain percentages by weight and all the oxygen present being substantially in the form of metal oxide. This control rod is characterized by its high strength and resistance to creep at elevated temperatures.

  13. Software Unit Testing during the Development of Digital Reactor Protection System of HTR-PM

    International Nuclear Information System (INIS)

    Guo Chao; Xiong Huasheng; Li Duo; Zhou Shuqiao; Li Jianghai

    2014-01-01

    Reactor Protection System (RPS) of High Temperature Gas-Cooled Reactor - Pebble bed Module (HTR-PM) is the first digital RPS designed and to be operated in the Nuclear Power Plant (NPP) of China, and its development process has receives a lot of concerns around the world. As a 1E-level safety system, the RPS has to be designed and developed following a series of nuclear laws and technical disciplines including software verification and validation (software V&V). Software V&V process demonstrates whether all stages during the software development are performed correctly, completely, accurately, and consistently, and the results of each stage are testable. Software testing is one of the most significant and time-consuming effort during software V&V. In this paper, we give a comprehensive introduction to the software unit testing during the development of RPS in HTR-PM. We first introduce the objective of the testing for our project in the aspects of static testing, black-box testing, and white-box testing. Then the testing techniques, including static testing and dynamic testing, are explained, and the testing strategy we employed is also introduced. We then introduce the principles of three kinds of coverage criteria we used including statement coverage, branch coverage, and the modified condition/decision coverage. As a 1E-level safety software, testing coverage needs to be up to 100% mandatorily. Then we talk the details of safety software testing during software development in HTR-PM, including the organization, methods and tools, testing stages, and testing report. The test result and experiences are shared and finally we draw a conclusion for the unit testing process. The introduction of this paper can contribute to improve the process of unit testing and software development for other digital instrumentation and control systems in NPPs. (author)

  14. Life-critical digital flight control systems

    Science.gov (United States)

    Mcwha, James

    1990-01-01

    Digital autopilot systems were first used on commercial airplanes in the late 1970s. The A-320 airplane was the first air transport airplane with a fly-by-wire primary flight control system. On the 767-X (777) airplane Boeing will install all fly-by-wire flight controls. Activities related to safety, industry status and program phases are discussed.

  15. Stress analysis of magnetically controlled reactor

    Directory of Open Access Journals (Sweden)

    Ben Tong

    2017-05-01

    Full Text Available To provide technique references for vibration reduction of magnetically controlled reactors (MCRs, stress, which is the inherent reason of vibration and noise, should be investigated. Stresses in reactor cores are produced due to the magnetostriction deformation of silicon steel and electromagnetic force between the core discs. So far, stress analysis on reactor cores was based on one-way coupled numerical method, which did not consider the influence of the stress on magnetic properties of the core material. Thus, multi-group magnetization and magnetostriction characteristics curves of silicon steel under different tensile stresses are measured firstly to support the computation. From the experiment results, it can be seen that magnetic properties of silicon steel change with stress. Then an electromagneto-mechanical two-way coupled numerical model for MCRs considering magnetostrictive effect and electromagnetic force effect is proposed. Stress distribution of MCR cores under the combination excitation of the sinusoidal wave current and different direct currents are calculated. From the computed results, it can be seen that a larger direct current has greater influence on MCRs vibration, which provides a theory basis for further analysis of vibration and noise reduction.

  16. Instrumentation and control strategies for an integral pressurized water reactor

    Directory of Open Access Journals (Sweden)

    Belle R. Upadhyaya

    2015-03-01

    Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

  17. Experimental evaluation of reactivity constraints for the closed-loop control of reactor power

    International Nuclear Information System (INIS)

    Bernard, J.A.; Lanning, D.D.; Ray, A.

    1984-01-01

    General principles for the closed-loop, digital control of reactor power have been identified, quantitatively enumerated, and experimentally demonstrated on the 5 MWt Research Reactor, MITR-II. The basic concept is to restrict the net reactivity so that it is always possible to make the reactor period infinite at the desired termination point of a transient by reversing the direction of motion of whatever control mechanism is associated with the controller. This capability is formally referred to as ''feasibility of control''. A series of ten experiments have been conducted over a period of eighteen months to demonstrate the efficacy of this property for the automatic control of reactor power. It has been shown that a controller which possesses this property is capable of both raising and lowering power in a safe, efficient manner while using a control rod of varying differential worth, that the reactivity constraints are a sufficient condition for the automatic control of reactor power, and that the use of a controller based on reactivity constraints can prevent overshoots either due to attempts to control a transient with a control rod of insufficient differential worth or due to failure to properly estimate when to commence rod insertion. Details of several of the more significant tests are presented together with a discussion of the rationale for the development of closed-loop control in large commercial power systems. Specific consideration is given to the motivation for designing a controller based on feasibility of control and the associated licensing issues

  18. Control rod for a nuclear reactor

    International Nuclear Information System (INIS)

    Roman, W.G.; Sutton, H.G. Jr.

    1979-01-01

    A control rod assembly for a nuclear reactor is disclosed having a remotely disengageable coupling between the control rod and the control rod drive shaft. The coupling is actuated by first lowering then raising the drive shaft. The described motion causes axial repositioning of a pin in a grooved rotatable cylinder, each being attached to different parts of the drive shaft which are axially movable relative to each other. In one embodiment, the relative axial motion of the parts of the drive shaft is used either to couple or to uncouple the connection by forcing resilient members attached to the drive shaft into or out of shouldered engagement, respectively, with an indentation formed in the control rod

  19. Control rod for a nuclear reactor

    International Nuclear Information System (INIS)

    Roman, W.G.; Sutton, H.G. Jr.

    1976-01-01

    A control rod assembly for a nuclear reactor is disclosed having a remotely disengageable coupling between the control rod and the control rod drive shaft. The coupling is actuated by first lowering then raising the drive shaft. The described motion causes axial repositioning of a pin in a grooved rotatable cylinder, each being attached to different parts of the drive shaft which are axially movable relative to each other. In one embodiment, the relative axial motion of the parts of the drive shaft is used either to couple or to uncouple the connection by forcing resilent members attached to the drive shaft into or out of shouldered engagement, respectively, with an indentation formed in the control rod

  20. CONTROLLED SHUNT REACTORS FOR ELECTRIC NETWORKS

    Directory of Open Access Journals (Sweden)

    Dolgopolov A.G.,

    2011-12-01

    Full Text Available The article presents results of the research and design of controlled shunt alternative current reactors (CSR. The analysis of domestic and foreign experience of the development and deployment of CSR is performed, the effectiveness of their applications in power systems is assessed and results of the tests of samples CSR-220 kV and above are shown. Constructive features of CSR circuit are described; technical characteristics of the CSR-220, 500 kV are given. The prospects for widespread introduction of CSR for the control of power systems regimes are shown. The application of CSR in combination with other control devices such as FACTS allows, based on high-voltage lines of high capacity, creating controlled transmission lines of new generation, which corresponds to all necessary requirements with time-developing power systems and its associations.

  1. A digital joint controller for manipulators

    International Nuclear Information System (INIS)

    Holt, E.J.; Palmer, D.E.B.

    1993-01-01

    Nuclear Electric's hydraulic heavy duty manipulators are used at a number of Magnox Power Stations for a wide variety of tasks. In recent years there has been a trend towards the use of manipulators for tasks requiring increasing precision of tip positioning. In order to meet this requirement, a digital controller has been designed to replace the analogue controller board used in almost all manipulator control systems. The new controller allows the programming of a wide range of closed loop control algorithms. Position and drive signal data may be passed to and from the controller by digital means, allowing direct connection to a graphical display system and/or a computer executing a guidance algorithm. The hardware and software design are outlined and performance in the laboratory and the field is reported. (author)

  2. Computer simulation system of neural PID control on nuclear reactor

    International Nuclear Information System (INIS)

    Chen Yuzhong; Yang Kaijun; Shen Yongping

    2001-01-01

    Neural network proportional integral differential (PID) controller on nuclear reactor is designed, and the control process is simulated by computer. The simulation result show that neutral network PID controller can automatically adjust its parameter to ideal state, and good control result can be gotten in reactor control process

  3. Realtime control of biogas reactors. Technical report

    Energy Technology Data Exchange (ETDEWEB)

    Poulsen, Allan K.

    2010-12-15

    In this project several online methods were connected to a biogas pilot plant designed and built by Xergi A/S (Foulum, Denmark). The pilot plant was composed of two stainless steel tanks used as substrate storage and as digester, respectively. The total volume of the reactor tank was 300 L, the working volume 200 L and the headspace volume 100 L. The process temperature in the biogas reactor was maintained at 52 {+-} 0.5 deg. C during normal operating conditions. The biogas production was measured with a flow meter and a controller was used for automatic control of temperature, effluent removal, feeding and for data logging. A NIRS (near infrared spectrometer) was connected to a recurrent loop measuring on the slurry while a {mu}-GC (micro gas chromatograph) and a MIMS (membrane inlet mass spectrometer) enabled online measurements of the gas phase composition. During the project period three monitoring campaigns were accomplished. The loading rate of the biogas reactor was increased stepwise during the periods while the process was monitored. In the first two campaigns the load was increased by increasing the mass of organic material added to the reactor each day. However, this increasing amount changed the retention time in the reactor and in order to keep the retention time constant an increasing amount of inhibitor of the microbial process was instead added in the third campaign and as such maintaining a constant organic load mass added to the reactor. The effect is similar to an increase in process load, while keeping the load of organic material and hence retention time constant. Methods have been developed for the following online technologies and each technology has been evaluated with regard to future use as a tool for biogas process monitoring: 1) {mu}-GC was able to quantitative monitor important gas phase parameters in a reliable, fast and low-maintenance way. 2) MIMS was able to quantitative monitor gas phase composition in a reliable and fast manner

  4. Digital transformation. With the increasing digitalisation of reactors, the nuclear sector must take the risk of cyber attacks into account. Atom in front of cyber crime

    International Nuclear Information System (INIS)

    Dupin, Ludovic

    2017-01-01

    After having evoked a science-fictional scenario of a cyber attack of a nuclear reactor, but also recent and actual, and sometimes successful cyber attacks against reactor control systems or uranium enrichment centrifuges, this article notices that authorities and bodies in charge of nuclear activities have become aware of this threat only for a short time, and that the threat is increased because of the increasing role of digital compounds in recent reactors. Therefore, a definition of good practices is emerging. In a brief interview, a manager of Assystem outlines that data theft is the main risk

  5. Method and practice on safety software verification and validation for digital reactor protection system

    International Nuclear Information System (INIS)

    Li Duo; Zhang Liangju; Feng Junting

    2010-01-01

    The key issue arising from digitalization of reactor protection system for Nuclear Power Plant (NPP) is in essence, how to carry out Verification and Validation (V and V), to demonstrate and confirm the software is reliable enough to perform reactor safety functions. Among others the most important activity of software V and V process is unit testing. This paper discusses the basic concepts on safety software V and V and the appropriate technique for software unit testing, focusing on such aspects as how to ensure test completeness, how to establish test platform, how to develop test cases and how to carry out unit testing. The technique discussed herein was successfully used in the work of unit testing on safety software of a digital reactor protection system. (author)

  6. Digital control of electric drives

    CERN Document Server

    Koziol, R; Szklarski, L

    1992-01-01

    The electromechanical systems employed in different branches of industry are utilized most often as drives of working machines which must be fed with electric energy in a continuous, periodic or even discrete way. Some of these machines operate at constant speed, others require wide and varying energy control. In many designs the synchronous cooperation of several electric drives is required in addition to the desired dynamic properties. For these reasons the control of the cooperation and dynamics of electromechanical systems requires the use of computers.This book adopts an unusual approach

  7. Replacement of the complete control system of the NPP Oskarshamn 1 by digital distributed control system

    International Nuclear Information System (INIS)

    Berger, E.

    1998-01-01

    As part of an ongoing modernization program, the I and C system and the control room of Oskarshamn 1 will be upgraded by ABB using its 'Advant Power' range of digital, programmable process control system. Besides ensuring the higher level of safety that is demanded today, the new equipment provides the plant with an integrated system which will improve operator interaction with the plant and reduce the risk of human error. The newly installed DCS system will serve also as a platform for further improvements of the control room. This paper discusses Oskarshamn 1 exchange of the complete control system of a nuclear power plant, the technical solution and the time schedule. Oskarshamn 1 is the first nuclear power plant in Sweden. It is a boiling water reactor built between 1966 and 1971 by ABB ATOM in Sweden. According to the plant age the control system is relay-based, while instrumentation and analogue control is semiconductor-based. This makes maintenance expensive and even worse, makes extensions nearly impossible. According to the safety standards of the 1960s, there is no separation between safety and non safety control and no seismic qualification. To extend the life of this plant the owner has decided to improve the safety system as well as to replace the reactor protection system, the safety related control and the non safety related control by a state-of-the-art digital distributed control system from ABB. In March 1997, ABB got the order to replace the reactor protection system, the safety control system and to start the replacement of all control systems. The old control room has to be replaced by a new ergonomically design. Together with the exchange of the control system the safety features of the plant and the emergency power supply has to be extended. (author)

  8. Migration of Older to New Digital Control Systems in Nuclear Power Plant Main Control Rooms

    Energy Technology Data Exchange (ETDEWEB)

    Kovesdi, Casey Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States); Joe, Jeffrey Clark [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-04-01

    The United States (U.S.) Department of Energy (DOE) Office of Nuclear Energy (NE) has the primary mission to advance nuclear power by resolving socio-technical issues through research and development (R&D). One DOE-NE activity supporting this mission is the Light Water Reactor Sustainability (LWRS) program. LWRS has the overall objective to sustain the operation of existing commercial nuclear power plants (NPPs) through conducting R&D across multiple “pathways,” or R&D focus areas. The Advanced Instrumentation, Information, and Control (II&C) Systems Technologies pathway conducts targeted R&D to address aging and reliability concerns with the legacy instrumentation and control (I&C) and related information systems in operating U.S. NPPs. This work involves (1) ensuring that legacy analog II&C systems are not life-limiting issues for the LWR fleet, and (2) implementing digital II&C technology in a manner that enables broad innovation and business improvement in the NPP operating model. Under the LWRS Advanced II&C pathway, Human Factors experts at Idaho National Laboratory (INL) have been conducting R&D in support of NPP main control room (MCR) modernization activities. Work in prior years has focused on migrating analog I&C systems to new digital I&C systems (). In fiscal year 2016 (FY16), one new focus area for this research is migrating older digital I&C systems to new and advanced digital I&C systems. This report summarizes a plan for conducting a digital-to-digital migration of a legacy digital I&C system to a new digital I&C system in support of control room modernization activities.

  9. Upgrading the Siemens Argonaut Reactor Graz with a digital monitoring system

    International Nuclear Information System (INIS)

    Froehlich, O.; Ninaus, W.

    1999-01-01

    This paper presents a modern design of a reactor monitoring system (MS) which was developed for a research reactor. This MS is using digital concepts, and it is more flexible than an analog MS, it co-operates better with the user, and it is a very helpful tool for a training-reactor in an university environment. The heart of the system is a process computer, and it was possible to access all important signals and functions of the original nuclear instrumentation by additional hardware. The monitoring software was written in C for the platform '32Bit-DOS-protected-mode' and shows on several high-resolution screen pages all the collected signals and the working conditions of the reactor. Moreover, all signals which are recorded on the random access memory can be saved to the hard disk of the computer and may thereby be used offline as well.(author)

  10. Overview of the US program of controls for advanced reactors

    Energy Technology Data Exchange (ETDEWEB)

    White, J.D.; Sackett, J.I.; Monson, R.; Lindsay, R.W.; Carroll, D.G.

    1989-01-01

    An automated control system can incorporate control goals and strategies, assessment of present and future plant status, diagnostic evaluation and maintenance planning, and signal and command validation. It has not been feasible to employ these capabilities in conventional hard-wired, analog, control systems. Recent advances in computer-based digital data acquisition systems, process controllers, fiber-optic signal transmission artificial intelligence tools and methods, and small inexpensive, fast, large-capacity computers---with both numeric and symbolic capabilities---have provided many of the necessary ingredients for developing large, practical automated control systems. Furthermore, recent reactor designs which provide strong passive responses to operational upsets or accidents afford good opportunities to apply these advances in control technology. This paper presents an overall US national perspective for advanced controls research and development. The goals of high reliability, low operating cost and simple operation are described. The staged approach from conceptualization through implementation is discussed. Then the paper describes the work being done by ORNL, ANL and GE. The relationship of this work to the US commercial industry is also discussed.

  11. Overview of the US program of controls for advanced reactors

    International Nuclear Information System (INIS)

    White, J.D.; Sackett, J.I.; Monson, R.; Lindsay, R.W.; Carroll, D.G.

    1989-01-01

    An automated control system can incorporate control goals and strategies, assessment of present and future plant status, diagnostic evaluation and maintenance planning, and signal and command validation. It has not been feasible to employ these capabilities in conventional hard-wired, analog, control systems. Recent advances in computer-based digital data acquisition systems, process controllers, fiber-optic signal transmission artificial intelligence tools and methods, and small inexpensive, fast, large-capacity computers---with both numeric and symbolic capabilities---have provided many of the necessary ingredients for developing large, practical automated control systems. Furthermore, recent reactor designs which provide strong passive responses to operational upsets or accidents afford good opportunities to apply these advances in control technology. This paper presents an overall US national perspective for advanced controls research and development. The goals of high reliability, low operating cost and simple operation are described. The staged approach from conceptualization through implementation is discussed. Then the paper describes the work being done by ORNL, ANL and GE. The relationship of this work to the US commercial industry is also discussed

  12. Design and Implementation of a Fuzzy Controller for a TRIGA Mark III Reactor

    Directory of Open Access Journals (Sweden)

    Tonatiuh Rivero-Gutiérrez

    2012-01-01

    Full Text Available The design and testing of a fuzzy rule based controller to regulate the power of a TRIGA Mark III research nuclear reactor are presented. The design does not require the current exact parameters of the point kinetic equations of the reactor. Instead, from a qualitative analysis of the actions taken by the operators during the reactor’s operation, a set of control rules is derived. The rules cover the operation of the reactor from low levels of about dozens of watts up to its full power level of one megawatt. The controller is able to increase power from different initial values to a wide range of desired levels, maintaining constant levels for long periods of time. The controller’s output is the external reactivity, which is further converted to a control rod incremental movement. The fuzzy controller is implemented on the reactor’s digital operating console, and the results of a series of experiments are discussed.

  13. Neural Network Controller for the Pressurized Water Reactor Power Control

    International Nuclear Information System (INIS)

    Haggag, S.S.; Kotb, S.A.

    2017-01-01

    Although there have been some severe nuclear accidents such as Three Mile Island (USA), Chernobyl (Ukraine) and Fukushima (Japan), nuclear fission energy is still a source of clean energy that can substitute fossil fuels in a centralized way and in a great amount with commercial availability and economic competitiveness. Since the pressurized water reactor (PWR) is the most widely used nuclear fission reactor, it is safe, stable and efficient operation is meaningful to the current rebirth of the nuclear fission energy industry. Power-level regulation is an important technique which can deeply affect the operation stability and efficiency of PWRs (Pressurized Water Reactors ). This paper presents the effect of utilizing the Neural Network controller methodology in the power control model of the PWR. The Neural Network Controller was tested on a PWR model using the Matlab Simulink Interface. Two case studies were performed on the model using both the Neural Network method and the traditional rod speed program for controlling the nuclear power plant variables. The proposed controller presents a higher performance than that of the traditional rod speed program controller.

  14. Adaptive robust control of the EBR-II reactor

    International Nuclear Information System (INIS)

    Power, M.A.; Edwards, R.M.

    1996-01-01

    Simulation results are presented for an adaptive H ∞ controller, a fixed H ∞ controller, and a classical controller. The controllers are applied to a simulation of the Experimental Breeder Reactor II primary system. The controllers are tested for the best robustness and performance by step-changing the demanded reactor power and by varying the combined uncertainty in initial reactor power and control rod worth. The adaptive H ∞ controller shows the fastest settling time, fastest rise time and smallest peak overshoot when compared to the fixed H ∞ and classical controllers. This makes for a superior and more robust controller

  15. NGNP Reactor Coolant Chemistry Control Study

    International Nuclear Information System (INIS)

    Castle, Brian

    2010-01-01

    The main focus of this paper is to identify the most desirable ranges of impurity levels in the primary coolant to optimize component life in the primary circuit of the Next Generation Nuclear Plant (NGNP), which will either be a prismatic block or pebble bed reactor. The 'NGNP Reactor Coolant Chemistry Control Study' has identified the necessary components and the approach used to control the coolant chemistry in high-temperature gas-cooled reactors (HTGR) for eventual use in the Next Generation Nuclear Plant (NGNP). The focus of this study is to determine what technical developments have taken place since previous helium purification systems (HPSs) were operated in previous HTGRs. Since previous HTGR operational experience, a considerable number of corrosion studies have been performed on various high temperature materials in helium environments with various chemistries at elevated temperatures. This newer work has been reviewed and incorporated into this corrosion study to determine how this new information impacts the design requirements for modern HPSs. Research shows that there are ratios of key impurities that determine whether or not a protective oxide layer is formed or if a deleterious relationship is formed between the material and the coolant. These and other influencing factors are taken into consideration when considering the design requirements for the HPS for the NGNP. In addition to the metallic components in the primary circuit, there are other important material considerations that greatly impact the chemistry requirements for the helium coolant, which are the graphite core support structures. While the metallic components need a small amount of moisture in the coolant to form a protective oxide layer, the graphite needs a very dry environment to prevent degradation. Recommendations and conclusion are made that identify important HPS parameters that should be considered in the design of the HPS for the NGNP. Following the coolant chemistry

  16. Distributed digital control of accelerators

    International Nuclear Information System (INIS)

    Crowley-Milling, M.

    1984-01-01

    The special properties, requirements and limitations of distributed control systems are described. The main advantages of distributed over centralized systems can be summarized: (a) For geographically distributed systems, complete systems can be operated locally even when links to the centre and the central system are unavailable. Similarly, parts of the equipment can be tested and commissioned separately during the constructon period; (b) Distribution of tasks between a number of computers gives the possibility of parallel processing with a corresponding gain in speed and reduction in response time; (c) Computer configurations and operating systems can be tailored to suit the application. New requirements can be catered for by the addition of one or more computers to the network without disturbance of the existing system; (d) Local surveillance and testing can reduce the load at higher levels which only need to be informed if anything goes wrong. Diagnosis of faults can be easier, despite the increase in number of units, as they can be arranged to test each other. Maintenance can be easier in a modular system, by exchange of modules, and it is easier to provide redundancy in vital parts of the system; (e) Different groups can work on different parts of the system with the minimum of interference; (f) The needs for bandwidth in the communications system for a large machine are lower for a decentralized than for a centralizzed system; (g) A distributed system can have economic advantages. It may require more hardware, which is becoming cheaper, but the software, which, if anything, is becoming more expensive, is generally simpler, and replicated

  17. Control theory of digitally networked dynamic systems

    CERN Document Server

    Lunze, Jan

    2013-01-01

    The book gives an introduction to networked control systems and describes new modeling paradigms, analysis methods for event-driven, digitally networked systems, and design methods for distributed estimation and control. Networked model predictive control is developed as a means to tolerate time delays and packet loss brought about by the communication network. In event-based control the traditional periodic sampling is replaced by state-dependent triggering schemes. Novel methods for multi-agent systems ensure complete or clustered synchrony of agents with identical or with individual dynamic

  18. Active disturbance rejection controller for chemical reactor

    International Nuclear Information System (INIS)

    Both, Roxana; Dulf, Eva H.; Muresan, Cristina I.

    2015-01-01

    In the petrochemical industry, the synthesis of 2 ethyl-hexanol-oxo-alcohols (plasticizers alcohol) is of high importance, being achieved through hydrogenation of 2 ethyl-hexenal inside catalytic trickle bed three-phase reactors. For this type of processes the use of advanced control strategies is suitable due to their nonlinear behavior and extreme sensitivity to load changes and other disturbances. Due to the complexity of the mathematical model an approach was to use a simple linear model of the process in combination with an advanced control algorithm which takes into account the model uncertainties, the disturbances and command signal limitations like robust control. However the resulting controller is complex, involving cost effective hardware. This paper proposes a simple integer-order control scheme using a linear model of the process, based on active disturbance rejection method. By treating the model dynamics as a common disturbance and actively rejecting it, active disturbance rejection control (ADRC) can achieve the desired response. Simulation results are provided to demonstrate the effectiveness of the proposed method

  19. Control of SHARON reactor for autotrophic nitrogen removal in two-reactor configuration

    DEFF Research Database (Denmark)

    Valverde Perez, Borja; Mauricio Iglesias, Miguel; Sin, Gürkan

    2012-01-01

    With the perspective of investigating a suitable control design for autotrophic nitrogen removal, this work explores the control design for a SHARON reactor. With this aim, a full model is developed, including the pH dependency, in order to simulate the reactor and determine the optimal operating...

  20. Fixed-bed Reactor Dynamics and Control - A Review

    DEFF Research Database (Denmark)

    Jørgensen, S. B.

    1986-01-01

    The industrial diversity of fixed bed reactors offers a challenging and relevant set of control problems. These intricate problems arise due to the rather complex dynamics of fixed bed reactors and to the complexity of actual reactor configurations. Many of these control problems are nonlinear...... and multi-variable. During the last decade fixed bed reactor control strategies have been proposed and investigated experimentally. This paper reviews research on these complex control problems with an emphasis upon solutions which have been demon-strated to work in the laboratory and hold promise...

  1. Neutron density optimal control of A-1 reactor analoque model

    International Nuclear Information System (INIS)

    Grof, V.

    1975-01-01

    Two applications are described of the optimal control of a reactor analog model. Both cases consider the control of neutron density. Control loops containing the on-line controlled process, the reactor of the first Czechoslovak nuclear power plant A-1, are simulated on an analog computer. Two versions of the optimal control algorithm are derived using modern control theory (Pontryagin's maximum principle, the calculus of variations, and Kalman's estimation theory), the minimum time performance index, and the quadratic performance index. The results of the optimal control analysis are compared with the A-1 reactor conventional control. (author)

  2. Dynamics and Control of Chemical Reactors-Selectively Surveyed

    DEFF Research Database (Denmark)

    Jørgensen, S. B.; Jensen, N.

    1989-01-01

    in industry, many reactor control problems are still left unsolved or only partly solved using open loop strategies where disturbance rejection and model inaccuracies have to be handled through manual reactor control and feedback control of raw material preprocessing and product purification operations...

  3. Distributed and recoverable digital control system

    Science.gov (United States)

    Stange, Kent (Inventor); Hess, Richard (Inventor); Kelley, Gerald B (Inventor); Rogers, Randy (Inventor)

    2010-01-01

    A real-time multi-tasking digital control system with rapid recovery capability is disclosed. The control system includes a plurality of computing units comprising a plurality of redundant processing units, with each of the processing units configured to generate one or more redundant control commands. One or more internal monitors are employed for detecting data errors in the control commands. One or more recovery triggers are provided for initiating rapid recovery of a processing unit if data errors are detected. The control system also includes a plurality of actuator control units each in operative communication with the computing units. The actuator control units are configured to initiate a rapid recovery if data errors are detected in one or more of the processing units. A plurality of smart actuators communicates with the actuator control units, and a plurality of redundant sensors communicates with the computing units.

  4. Liquid-poison type power controlling device for nuclear reactor

    International Nuclear Information System (INIS)

    Horiuchi, Tetsuo; Yamanari, Shozo; Sugisaki, Toshihiko; Goto, Hiroshi.

    1981-01-01

    Purpose: To improve the safety and the operability of a nuclear reactor by adjusting the density of liquid poison. Constitution: The thermal expansion follow-up failure between cladding and a pellet upon abrupt and local variations of the power is avoided by adjusting the density of liquid poison during ordinary operation in combination with a high density liquid poison tank and a filter and smoothly controlling the reactor power through a pipe installed in the reactor core. The high density liquid poison is abruptly charged in to the reactor core under relatively low pressure through the tube installed in the reactor core at the time of control rod insertion failure in an accident, thereby effectively shutting down the reactor and improving the safety and the operability of the reactor. (Yoshihara, H.)

  5. The control of emissions from nuclear power reactors in Canada

    International Nuclear Information System (INIS)

    Gorman, D.J.; Neil, B.C.J.; Chatterjee, R.M.

    1988-01-01

    Nuclear power reactors in Canada are of the CANDU pressurised heavy water design. These are located in the provinces of Ontario, Quebec, and New Brunswick. Most of the nuclear generating capacity is in the province of Ontario which has 16 commissioned reactors with a total capacity of 11,500 MWe. There are four reactors under construction with an additional capacity of 3400 MWe. Nuclear power currently accounts for approximately 50% of the electrical power generation of Ontario. Regulation of the reactors is a Federal Government responsibility administered by the Atomic Energy Control Board (AECB) which licenses the reactors and sets occupational and public dose limits

  6. Optimal control rod programs in power reactors

    International Nuclear Information System (INIS)

    Fadilah, S.M.; Lewins, J.

    1975-01-01

    Control rod programming is investigated with respect to optimising the power peaking factor and hence the utilisation of a nuclear reactor. A simplified diffusion model, initially with a finite number of regions, in cylindrical geometry, is used to enable optimal trajectories to be completely synthesised. The average discharge burnup problem is posed both as an external and as an internal optimisation. The connection between optimum power shape and the maximisation of the average discharge burnup is explored in a wider context. It is shown that optimum trajectories combine an initial singular solution of the Haling type with a terminal bang-bang solution. An extension to a higher number of regions and, on passing to the limit, to a diffusion model, provides an alternative proof of Haling's principle without the restriction to monotonic reactivity decrease with burnup. Numerical results in the two-region model are given to show the general scope of optimisation available. (author)

  7. Adaptive nonlinear control for a research reactor

    International Nuclear Information System (INIS)

    Benitez R, J.S.

    1994-01-01

    Linearization by feedback of states is based on the idea of transform the nonlinear dynamic equation of a system in a linear form. This linear behavior can be achieve well in a complete way (input state) or in partial way (input output). This can be applied to systems of single or multiple inputs, and can be used to solve problems of stabilization and tracking of references trajectories. Comparing this method with conventional ones, linearization by feedback of states is exact in certain region of the space of state, instead of linear approximations of the equations in a certain point of the operation. In the presence of parametric uncertainties in the model of the system, the introduction of adaptive schemes provide a type toughness to the control system by nonlinear feedback, which gives as result the eventual cancellation of the nonlinear terms in the dynamic relationship between the output and the input of the auxiliary control. In the same way, it has been presented the design of a nonlinearizing control for the non lineal model of a TRIGA Mark III type reactor, with the aim of tracking a predetermined power profile. The asymptotic tracking of such profile is, at the present moment, in the stage of verification by computerized simulation the relative easiness in the design of auxiliary variable of control, as well as the decoupling action of the output variable, make very attractive the utilization of the method herein presented. (Author)

  8. Functional issues and environmental qualification of digital protection systems of advanced light-water nuclear reactors

    International Nuclear Information System (INIS)

    Korsah, K.; Clark, R.L.; Wood, R.T.

    1994-04-01

    Issues of obsolescence and lack of infrastructural support in (analog) spare parts, coupled with the potential benefits of digital systems, are driving the nuclear industry to retrofit analog instrumentation and control (I ampersand C) systems with digital and microprocessor-based systems. While these technologies have several advantages, their application to safety-related systems in nuclear power plants raises key issues relating to the systems' environmental qualification and functional reliability. To bound the problem of new I ampersand C system functionality and qualification, the authors focused this study on protection systems proposed for use in ALWRs. Specifically, both functional and environmental qualification issues for ALWR protection system I ampersand C were addressed by developing an environmental, functional, and aging data template for a protection division of each proposed ALWR design. By using information provided by manufacturers, environmental conditions and stressors to which I ampersand C equipment in reactor protection divisions may be subjected were identified. The resulting data were then compared to a similar template for an instrument string typically found in an analog protection division of a present-day nuclear power plant. The authors also identified fiber-optic transmission systems as technologies that are relatively new to the nuclear power plant environment and examined the failure modes and age-related degradation mechanisms of fiber-optic components and systems. One reason for the exercise of caution in the introduction of software into safety-critical systems is the potential for common-cause failure due to the software. This study, however, approaches the functionality problem from a systems point of view. System malfunction scenarios are postulated to illustrate the fact that, when dealing with the performance of the overall integrated system, the real issues are functionality and fault tolerance, not hardware vs. software

  9. Functional issues and environmental qualification of digital protection systems of advanced light-water nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Korsah, K.; Clark, R.L.; Wood, R.T. [Oak Ridge National Lab., TN (United States)

    1994-04-01

    Issues of obsolescence and lack of infrastructural support in (analog) spare parts, coupled with the potential benefits of digital systems, are driving the nuclear industry to retrofit analog instrumentation and control (I&C) systems with digital and microprocessor-based systems. While these technologies have several advantages, their application to safety-related systems in nuclear power plants raises key issues relating to the systems` environmental qualification and functional reliability. To bound the problem of new I&C system functionality and qualification, the authors focused this study on protection systems proposed for use in ALWRs. Specifically, both functional and environmental qualification issues for ALWR protection system I&C were addressed by developing an environmental, functional, and aging data template for a protection division of each proposed ALWR design. By using information provided by manufacturers, environmental conditions and stressors to which I&C equipment in reactor protection divisions may be subjected were identified. The resulting data were then compared to a similar template for an instrument string typically found in an analog protection division of a present-day nuclear power plant. The authors also identified fiber-optic transmission systems as technologies that are relatively new to the nuclear power plant environment and examined the failure modes and age-related degradation mechanisms of fiber-optic components and systems. One reason for the exercise of caution in the introduction of software into safety-critical systems is the potential for common-cause failure due to the software. This study, however, approaches the functionality problem from a systems point of view. System malfunction scenarios are postulated to illustrate the fact that, when dealing with the performance of the overall integrated system, the real issues are functionality and fault tolerance, not hardware vs. software.

  10. Stability of digital feedback control systems

    Directory of Open Access Journals (Sweden)

    Larkin Eugene

    2018-01-01

    Lag time characteristics are used for investigation of stability of linear systems. Digital PID controller is divided onto linear part, which is realized with a soft and pure lag unit, which is realized with both hardware and software. With use notions amplitude and phase margins, condition for stability of system functioning are obtained. Theoretical results are confirm with computer experiment carried out on the third-order system.

  11. Polynomial Digital Control of a Series Equal Liquid Tanks

    Directory of Open Access Journals (Sweden)

    Bobála Vladimír

    2016-01-01

    Full Text Available Time-delays are mainly caused by the time required to transport mass, energy or information, but they can also be caused by processing time or accumulation. Typical examples of such processes are e.g. pumps, liquid storing tanks, distillation columns or some types of chemical reactors. In many cases time-delay is caused by the effect produced by the accumulation of a large number of low-order systems. Several industrial processes have the time-delay effect produced by the accumulation of a great number of low-order systems with the identical dynamic. The dynamic behavior of series these low-order systems is expressed by high-order system. One of possibilities of control of such processes is their approximation by low-order model with time-delay. The paper is focused on the design of the digital polynomial control of a set of equal liquid cylinder atmospheric tanks. The designed control algorithms are realized using the digital Smith Predictor (SP based on polynomial approach – by minimization of the Linear Quadratic (LQ criterion. The LQ criterion was combined with pole assignment.

  12. Control of SHARON reactor for autotrophic nitrogen removal in two-reactor configuration

    DEFF Research Database (Denmark)

    Valverde Perez, Borja; Mauricio Iglesias, Miguel; Sin, Gürkan

    2012-01-01

    With the perspective of investigating a suitable control design for autotrophic nitrogen removal, this work explores the control design for a SHARON reactor. With this aim, a full model is developed, including the pH dependency, in order to simulate the reactor and determine the optimal operating...... conditions. Then, the screening of controlled variables and pairing is carried out by an assessment of the effect of the disturbances based on the closed loop disturbance gain plots. Two controlled structures are obtained and benchmarked by their capacity to reject the disturbances before the Anammox reactor....

  13. APPLICATION OF MODEL PREDICTIVE CONTROL TO BATCH POLYMERIZATION REACTOR

    Directory of Open Access Journals (Sweden)

    N.M. Ghasem

    2006-06-01

    Full Text Available The absence of a stable operational state in polymerization reactors that operates in batches is factor that determine the need of a special control system. In this study, advanced control methodology is implemented for controlling the operation of a batch polymerization reactor for polystyrene production utilizingmodel predictive control. By utilizing a model of the polymerization process, the necessary operational conditions were determined for producing the polymer within the desired characteristics. The maincontrol objective is to bring the reactor temperature to its target temperature as rapidly as possible with minimal temperature overshoot. Control performance for the proposed method is encouraging. It has been observed that temperature overshoot can be minimized by the proposed method with the use of both reactor and jacket energy balance for reactor temperature control.

  14. Technical quality control - constancy controls for digital mammography systems

    International Nuclear Information System (INIS)

    Pedersen, K.; Landmark, I.D.; Bredholt, K.; Hauge, I.H.R.

    2009-04-01

    To ensure the quality of mammographic images, so-called constancy control tests are performed frequently. The report contains a programme for constancy control of digital mammography systems, encompassing the mammography unit, computed radiography (CR) systems, viewing conditions and displays, printers, and procedures for data collection for patient dose calculations. (Author)

  15. Formal verification and validation of the safety-critical software in a digital reactor protection system

    International Nuclear Information System (INIS)

    Kwon, K. C.; Park, G. Y.

    2006-01-01

    This paper describes the Verification and Validation (V and V) activities for the safety-critical software in a Digital Reactor Protection System (DRPS) that is being developed through the Korea nuclear instrumentation and control system project. The main activities of the DRPS V and V process are a preparation of the software planning documentation, a verification of the software according to the software life cycle, a software safety analysis and a software configuration management. The verification works for the Software Requirement Specification (SRS) of the DRPS consist of a technical evaluation, a licensing suitability evaluation, a inspection and traceability analysis, a formal verification, and preparing a test plan and procedure. Especially, the SRS is specified by the formal specification method in the development phase, and the formal SRS is verified by a formal verification method. Through these activities, we believe we can achieve the functionality, performance, reliability, and safety that are the major V and V objectives of the nuclear safety-critical software in a DRPS. (authors)

  16. Digital computer control of servomotor angular position | Mullisa ...

    African Journals Online (AJOL)

    The paper discussess the design and simulation methodology of digital control systems for the benefit of the interested practicing engineer. A lead-type digital controller for a 2nd order system and a leadlag type digital controller for a 3rd order system are designed. The simulations show that the design methods are ...

  17. Research and development on the application of advanced control technologies to advanced nuclear reactor systems: A US national perspective

    International Nuclear Information System (INIS)

    White, J.D.; Monson, L.R.; Carrol, D.G.; Dayal, Y.

    1989-01-01

    Control system designs for nuclear power plants are becoming more advanced through the use of digital technology and automation. This evolution is taking place because of: (1) the limitations in analog based control system performance and maintenance and availability and (2) the promise of significant improvement in plant operation and availability due to advances in digital and other control technologies. Digital retrofits of control systems in US nuclear plants are occurring now. Designs of control and protection systems for advanced LWRs are based on digital technology. The use of small inexpensive, fast, large-capacity computers in these designs is the first step of an evolutionary process described in this paper. Under the sponsorship of the US Department of Energy (DOE), Oak Ridge National Laboratory, Argonne National Laboratory, GE Nuclear Energy and several universities are performing research and development in the application of advances in control theory, software engineering, advanced computer architectures, artificial intelligence, and man-machine interface analysis to control system design. The target plant concept for the work described in this paper is the Power Reactor Inherently Safe Module reactor (PRISM), an advanced modular liquid metal reactor concept. This and other reactor designs which provide strong passive responses to operational upsets or accidents afford good opportunities to apply these advances in control technology. 18 refs., 5 figs

  18. Nuclear reactor power control system based on flexibility model

    International Nuclear Information System (INIS)

    Li Gang; Zhao Fuyu; Li Chong; Tai Yun

    2011-01-01

    Design the nuclear reactor power control system in this paper to cater to a nonlinear nuclear reactor. First, calculate linear power models at five power levels of the reactor as five local models and design controllers of the local models as local controllers. Every local controller consists of an optimal controller contrived by the toolbox of Optimal Controller Designer (OCD) and a proportion-integration-differentiation (PID) controller devised via Genetic Algorithm (GA) to set parameters of the PID controller. According to the local models and controllers, apply the principle of flexibility model developed in the paper to obtain the flexibility model and the flexibility controller at every power level. Second, the flexibility model and the flexibility controller at a level structure the power control system of this level. The set of the whole power control systems corresponding to global power levels is to approximately carry out the power control of the reactor. Finally, the nuclear reactor power control system is simulated. The simulation result shows that the idea of flexibility model is feasible and the nuclear reactor power control system is effective. (author)

  19. High Performance Low Cost Digitally Controlled Power Conversion Technology

    DEFF Research Database (Denmark)

    Jakobsen, Lars Tønnes

    2008-01-01

    Digital control of switch-mode power supplies and converters has within the last decade evolved from being an academic subject to an emerging market in the power electronics industry. This development has been pushed mainly by the computer industry that is looking towards digital power management...... the execution time of the software algorithm that realises the digital control law will constitute a considerable delay in the control loop. Digital signal controllers are powerful devices capable of performing arithmetic functions much faster than a microcontroller can. Digital signal controllers are well...... and an analogue to digital converter with a short sampling time. A digital self-oscillating modulator is proposed in the present thesis. The modulator is a free-running modulator which operates without an external carrier signal. Customised digital control solutions offers the best performance for non-isolated DC...

  20. Fully integrated analysis of reactor kinetics, thermalhydraulics and the reactor control system in the MAPLE-X10 research reactor

    International Nuclear Information System (INIS)

    Shim, S.Y.; Carlson, P.A.; Baxter, D.K.

    1992-01-01

    A prototype research reactor, designated MAPLE-X10 (Multipurpose Applied Physics Lattice Experimental - X 10MW), is currently being built at AECL's Chalk River Laboratories. The CATHENA (Canadian Algorithm for Thermalhydraulic Network Analysis) two-fluid code was used in the safety analysis of the reactor to determine the adequacy of core cooling during postulated reactivity and loss-of-forced-flow transients. The system responses to a postulated transient are predicted including the feedback between reactor kinetics, thermalhydrauilcs and the reactor control systems. This paper describes the MAPLE-X10 reactor and the modelling methodology used. Sample simulations of postulated loss-of-heat-sink and loss-of-regulation transients are presented. (author)

  1. Light Water Reactor Sustainability Program: Digital Technology Business Case Methodology Guide

    Energy Technology Data Exchange (ETDEWEB)

    Thomas, Ken [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lawrie, Sean [ScottMadden, Inc., Raleigh, NC (United States); Hart, Adam [ScottMadden, Inc., Raleigh, NC (United States); Vlahoplus, Chris [ScottMadden, Inc., Raleigh, NC (United States)

    2014-09-01

    The Department of Energy’s (DOE’s) Light Water Reactor Sustainability Program aims to develop and deploy technologies that will make the existing U.S. nuclear fleet more efficient and competitive. The program has developed a standard methodology for determining the impact of new technologies in order to assist nuclear power plant (NPP) operators in building sound business cases. The Advanced Instrumentation, Information, and Control (II&C) Systems Technologies Pathway is part of the DOE’s Light Water Reactor Sustainability (LWRS) Program. It conducts targeted research and development (R&D) to address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals: (1) to ensure that legacy analog II&C systems are not life-limiting issues for the LWR fleet and (2) to implement digital II&C technology in a manner that enables broad innovation and business improvement in the NPP operating model. Resolving long-term operational concerns with the II&C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation’s energy and environmental security. The II&C Pathway is conducting a series of pilot projects that enable the development and deployment of new II&C technologies in existing nuclear plants. Through the LWRS program, individual utilities and plants are able to participate in these projects or otherwise leverage the results of projects conducted at demonstration plants. Performance advantages of the new pilot project technologies are widely acknowledged, but it has proven difficult for utilities to derive business cases for justifying investment in these new capabilities. Lack of a business case is often cited by utilities as a barrier to pursuing wide-scale application of digital technologies to nuclear plant work activities. The decision to move forward with funding usually hinges on

  2. Design and adjustment on test bed of replacing subassembly machine control system for China experimental fast reactor

    International Nuclear Information System (INIS)

    Dong Shengguo; Ma Hongsheng; Zhao Lixia

    2008-01-01

    The present research concerns in the design and adjustment of replacing sub- assembly machine control system of China Experimental Fast Reactor. The design of replacing subassembly machine control system adopts some electric equipments, such as programmable controllers, digital DC drivers. The designed control system was adjusted on the test bed. The results indicate that the operation of the control system is steady and reliable, and designed control system can meet the needs of the design specification. (authors)

  3. Fusion reactor control study. Volume 3. Tandem mirror reactors. Final report

    International Nuclear Information System (INIS)

    Chang, F.R.; DeCanio, F.; Fisher, J.L.; Madden, P.A.

    1982-03-01

    A study of the control requirements of the Tandem Mirror Reactor concept is reported. The study describes the development of a control simulator that is based upon a spatially averaged physics code of the reactor concept. The simulator portrays the evolution of the plasma through the complete reactor operating cycle; it includes models of the control and measurement system, thus allowing the exploration of various strategies for reactor control. Startup, shutdown, and control during the quasi-steady-state power producing phase were explored. Configurations are described which use a variety of control effectors including modulation of the refueling rate, beam current, and electron cyclotron resonance heating. Multivariable design techniques were used to design the control laws and compensators for the feedback controllers and presume the practical measurement of only a subset of the plasma and machine variables. Performance of the various controllers is explored using the nonlinear control simulator. Derivative control strategies using new or developed sensors and effectors appropriate to a power reactor environment are postulated, based upon the results of the control configurations tested. Research and development requirements for these controls are delineated

  4. Integrated, digital experiment transient control and safety protection of an in-pile test

    International Nuclear Information System (INIS)

    Thomas, R.W.; Whitacre, R.F.; Klingler, W.B.

    1982-01-01

    The Sodium Loop Safety Facility experimental program has demonstrated that in-pile loop fuel failure transient tests can be digitally controlled and protected with reliability and precision. This was done in four nuclear experiments conducted in the Engineering Test Reactor operated by EG and G Idaho, Inc., at the Idaho National Engineering Laboratory. Loop sodium flow and reactor power transients can be programmed to sponsor requirements and verified prior to the test. Each controller has redundancy, which reduces the effect of single failures occurring during test transients. Feedback and reject criteria are included in the reactor power control. Timed sequencing integrates the initiation of the controllers, programmed safety set-points, and other experiment actions (e.g., planned scram). Off-line and on-line testing is included. Loss-of-flow, loss-of-piping-integrity, boiling-window, transient-overpower, and local fault tests have been successfully run using this system

  5. Regulatory Framework for Controlling the Research Reactor Decommissioning Project

    International Nuclear Information System (INIS)

    Melani, Ai; Chang, Soon Heung

    2009-01-01

    Decommissioning is one of important stages in construction and operation of research reactors. Currently, there are three research reactors operating in Indonesia. These reactors are operated by the National Nuclear Energy Agency (BATAN). The age of the three research reactors varies from 22 to 45 years since the reactors reached their first criticality. Regulatory control of the three reactors is conducted by the Nuclear Energy Regulatory Agency (BAPETEN). Controlling the reactors is carried out based on the Act No. 10/1997 on Nuclear Energy, Government Regulations and BAPETEN Chairman Decrees concerning the nuclear safety, security and safeguards. Nevertheless, BAPETEN still lack of the regulation, especially for controlling the decommissioning project. Therefore, in the near future BAPETEN has to prepare the regulations for decommissioning, particularly to anticipate the decommissioning of the oldest research reactors, which probably will be done in the next ten years. In this papers author give a list of regulations should be prepared by BAPETEN for the decommissioning stage of research reactor in Indonesia based on the international regulatory practice

  6. Feedwater control device for reactor pressure vessels

    International Nuclear Information System (INIS)

    Oonuma, Takeshi.

    1982-01-01

    Purpose: To prevent the generation of thermal stresses at the junction between a clean-up water pipe and a feedwater pipe. Constitution: Hot water containing impurities in a pressure vessel is caused to flow by a recycling pump through a heat exchanger, a cooler and a clean-up desalter and again by way of the heat exchanger into the feedwater pipe at the junction with the clean-up water pipe, where it is mixed with the feedwater passed by way of a feedwater heater and supplied to the pressure vessel. The feedwater temperature for the feedwater pipe and the set temperature for the clean-up water are compared with each other by using temperature sensors disposed to the feedwater pipe between the junction and the feedwater heater at the upstream of the junction. If the temperature difference is increased, for instance, upon transient state where the operation of the feedwater heater is not yet stabilized, the recycling pump is controlled to stop the supply of the clean-up water to the junction while flowing only the feedwater. This makes the temperature distribution uniform and prevents the generation of the thermal stresses at the junction, by which reactor safety can be improved. (Moriyama, K.)

  7. Hardware for digitally controlled scanned probe microscopes

    OpenAIRE

    Clark, S. M.; Baselt, D. R.; Spence, C. F.; Youngquist, M. G.; Baldeschwieler, J. D.

    1992-01-01

    The design and implementation of a flexible and modular digital control and data acquisition system for scanned probe microscopes (SPMs) is presented. The measured performance of the system shows it to be capable of 14-bit data acquisition at a 100-kHz rate and a full 18-bit output resolution resulting in less than 0.02-Å rms position noise while maintaining a scan range in excess of 1 µm in both the X and Y dimensions. This level of performance achieves the goal of making the noise of the mi...

  8. Digital signal processing in power electronics control circuits

    CERN Document Server

    Sozanski, Krzysztof

    2013-01-01

    Many digital control circuits in current literature are described using analog transmittance. This may not always be acceptable, especially if the sampling frequency and power transistor switching frequencies are close to the band of interest. Therefore, a digital circuit is considered as a digital controller rather than an analog circuit. This helps to avoid errors and instability in high frequency components. Digital Signal Processing in Power Electronics Control Circuits covers problems concerning the design and realization of digital control algorithms for power electronics circuits using

  9. CANDU Digital Control Computer upgrade options

    International Nuclear Information System (INIS)

    De Jong, M.S.; De Grosbois, J.; Qian, T.

    1997-01-01

    This paper reviews the evolution of Digital Control Computers (DCC) in CANDU power plants to the present day. Much of this evolution has been to meeting changing control or display requirements as well as the replacement of obsolete, or old and less reliable technology with better equipment that is now available. The current work at AECL and Canadian utilities to investigate DCC upgrade options, alternatives, and strategies are examined. The dependence of a particular upgrade strategy on the overall plant refurbishment plans are also discussed. Presently, the upgrade options range from replacement of individual obsolete system components, to replacement of the entire DCC hardware without changing the software, to complete replacement of the DCCs with a functionally equivalent system using new control computer equipment and software. Key issues, constraints and objectives associated with these DCC upgrade options are highlighted. (author)

  10. Lessons learned in digital upgrade projects digital control system implementation at US nuclear power stations

    International Nuclear Information System (INIS)

    Kelley, S.; Bolian, T. W.

    2006-01-01

    AREVA NP has gained significant experience during the past five years in digital upgrades at operating nuclear power stations in the US. Plants are seeking modernization with digital technology to address obsolescence, spare parts availability, vendor support, increasing age-related failures and diminished reliability. New systems offer improved reliability and functionality, and decreased maintenance requirements. Significant lessons learned have been identified relating to the areas of licensing, equipment qualification, software quality assurance and other topics specific to digital controls. Digital control systems have been installed in non safety-related control applications at many utilities within the last 15 years. There have also been a few replacements of small safety-related systems with digital technology. Digital control systems are proving to be reliable, accurate, and easy to maintain. Digital technology is gaining acceptance and momentum with both utilities and regulatory agencies based upon the successes of these installations. Also, new plants are being designed with integrated digital control systems. To support plant life extension and address obsolescence of critical components, utilities are beginning to install digital technology for primary safety-system replacement. AREVA NP analyzed operating experience and lessons learned from its own digital upgrade projects as well as industry-wide experience to identify key issues that should be considered when implementing digital controls in nuclear power stations

  11. Stop valve with automatic control and locking for nuclear reactors

    International Nuclear Information System (INIS)

    Chung, D.K.

    1980-01-01

    This invention generally concerns an automatic control and locking stop valve. Specifically it relates to the use of such a valve in a nuclear reactor of the type containing absorber elements supported by a fluid and intended for stopping the reactor in complete safety [fr

  12. Modelling and control design for SHARON/Anammox reactor sequence

    DEFF Research Database (Denmark)

    Valverde Perez, Borja; Mauricio Iglesias, Miguel; Sin, Gürkan

    2012-01-01

    With the perspective of investigating a suitable control design for autotrophic nitrogen removal, this work presents a complete model of the SHARON/Anammox reactor sequence. The dynamics of the reactor were explored pointing out the different scales of the rates in the system: slow microbial...

  13. Development of a nuclear reactor control system simulator using virtual instruments

    Energy Technology Data Exchange (ETDEWEB)

    Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Lameiras, Fernando Soares, E-mail: ajp@cdtn.b, E-mail: amir@cdtn.b, E-mail: fsl@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. This article describes a digital system being developed to simulate the behavior of the operating parameters using virtual instruments. The control objective is to bring the reactor power from its source level (mW) to a full power (kW). It is intended for education of basic reactor neutronic and thermohydraulic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron, control by rods, fuel and coolant temperatures, power, etc. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Centre - CDTN was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. The simulator system is being developed using the LabVIEW (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's) using electronic processor and visual interface in video monitor. The main purpose of the system is to provide training tools for instructors and students, allowing navigating by user-friendly operator interface and monitoring tendencies of the operational variables. It will be an interactive tool for training and teaching and could be used to predict the reactor behavior. Some scenarios are presented to demonstrate that it is possible to know the behavior of some variables from knowledge of input parameters. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility. (author)

  14. Development of a nuclear reactor control system simulator using virtual instruments

    International Nuclear Information System (INIS)

    Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Lameiras, Fernando Soares

    2011-01-01

    The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. This article describes a digital system being developed to simulate the behavior of the operating parameters using virtual instruments. The control objective is to bring the reactor power from its source level (mW) to a full power (kW). It is intended for education of basic reactor neutronic and thermohydraulic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron, control by rods, fuel and coolant temperatures, power, etc. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Centre - CDTN was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. The simulator system is being developed using the LabVIEW (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's) using electronic processor and visual interface in video monitor. The main purpose of the system is to provide training tools for instructors and students, allowing navigating by user-friendly operator interface and monitoring tendencies of the operational variables. It will be an interactive tool for training and teaching and could be used to predict the reactor behavior. Some scenarios are presented to demonstrate that it is possible to know the behavior of some variables from knowledge of input parameters. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility. (author)

  15. Control of a loop polymerization reactor using neural networks

    Directory of Open Access Journals (Sweden)

    M.P. Vega

    2000-12-01

    Full Text Available or multivariable non linear predictive control implementations, a hybrid-neural model (lumped model was successfully used for modeling a loop-tubular polymerization reactor (a lumped or distributed model, depending on recycle ratio. Bifurcation diagrams were computed in order to investigate the agreement between process and model, of paramount importance for model based controller implementation purposes. Performance was evaluated considering the nonlinear model predictive control of both a loop tubular reactor (lumped SISO problem and a tubular reactor (distributed MIMO problem.

  16. Reliability study: digital reactor protection system of Korean standard nuclear power plant

    International Nuclear Information System (INIS)

    Kang, H. G.; Jang, S. C.; Eom, H. S.; Jeong, H. S.

    2003-02-01

    Digital safety-critical systems which are now installed in Korean Standard Nuclear Power Plants (KSNPP) would be quantitatively evaluated in order to prove the safety. In this study, we quantify the safety of the digital reactor protection system in KSNPPs using PSA technology. This study also includes the detailed investigation of the target system operation. The Fault Tree (FT) models were constructed for 15 reactor trip parameters. For digital parts, because the operation data for the same type PWR was unavailable, we used the data provided by vendors. On the other hand, for the conventional analog/mechanical parts, we used experience data presented in KAERI/TR-2164/2002.The result of quantification shows that the system unavailability varies from 4.36E-5 to 8.96E-4 according to the trip parameter. Main contributor to the difference from the conventional analysis would be the difference in human failure probability estimation. Generally, the system unavailability depends on several important factors: Human failure probability, software failure probability, watchdog timer coverage, and common cause failure estimation

  17. Status of control assembly materials in Indian water reactors

    International Nuclear Information System (INIS)

    Date, V.G.; Kulkarni, P.G.

    2000-01-01

    India's present operating water cooled power reactors comprise boiling water reactors of Tarapur Atomic Power Station (TAPS) and pressurized heavy water reactors (PHWRs) at Kota (RAPS), Kalpakkam (MAPS), Narora (NAPS) and Kakrapara (KAPS). Boiling water reactors of TAPS use boron carbide control blades for control of power as well as for shut down (scram). PHWRs use boron steel and cobalt absorber rods for power control and Cd sandwiched shut off rods (primary shut down system) and liquid poison rods (secondary shut down system) for shut down. In TAPS, Gadolinium rods (burnable poison rods) are also incorporated in fuel assembly for flux flattening. Boron carbide control blades and Gadolinium rods for TAPS, cobalt absorber rods and shut down assemblies for PHWRs are fabricated indigenously. Considerable development work was carried out for evolving material specifications, component and assembly drawings, and fabrication processes. Details of various control and shut off assemblies being fabricated currently are highlighted in the paper. (author)

  18. Human error probability evaluation as part of reliability analysis of digital protection system of advanced pressurized water reactor - APR 1400

    International Nuclear Information System (INIS)

    Varde, P. V.; Lee, D. Y.; Han, J. B.

    2003-03-01

    A case of study on human reliability analysis has been performed as part of reliability analysis of digital protection system of the reactor automatically actuates the shutdown system of the reactor when demanded. However, the safety analysis takes credit for operator action as a diverse mean for tripping the reactor for, though a low probability, ATWS scenario. Based on the available information two cases, viz., human error in tripping the reactor and calibration error for instrumentations in protection system, have been analyzed. Wherever applicable a parametric study has also been performed

  19. Dynamics and control of molten-salt breeder reactor

    Directory of Open Access Journals (Sweden)

    Vikram Singh

    2017-08-01

    Full Text Available Preliminary results of the dynamic analysis of a two-fluid molten-salt breeder reactor (MSBR system are presented. Based on an earlier work on the preliminary dynamic model of the concept, the model presented here is nonlinear and has been revised to accurately reflect the design exemplified in ORNL-4528. A brief overview of the model followed by results from simulations performed to validate the model is presented. Simulations illustrate stable behavior of the reactor dynamics and temperature feedback effects to reactivity excursions. Stable and smooth changes at various nodal temperatures are also observed. Control strategies for molten-salt reactor operation are discussed, followed by an illustration of the open-loop load-following capability of the molten-salt breeder reactor system. It is observed that the molten-salt breeder reactor system exhibits “self-regulating” behavior, minimizing the need for external controller action for load-following maneuvers.

  20. Development of a system based in a digital signal processor (DSP) for a simulator of power regulation in a reactor: first stage

    International Nuclear Information System (INIS)

    Benitez R, J.S.; Perez C, B.

    2002-01-01

    The first stage of the development of a digital system based on a DSP is presented which forms part of an hybrid simulator for the power regulation in am model of the punctual kinetics of a TRIGA reactor type. The DSP performs the regulation, using a Mandami type algorithm of diffuse control. In the algorithm, the universe of the output variable is discretized for performing in an unique stage the aggregation functions and dis-diffusization. (Author)

  1. A Study of Digitally Controlled Flight Control Actuation

    Science.gov (United States)

    1983-07-01

    in 12 groups of three. CONTROL SERVO PROCESSOR PROCESSOR EhSV ACT.• ~INTERNAL , A Wm1R INC A, CHANNEL A1-12 MIL-STD-1553 B BUS A A C1.12 j CHANNEL 11...line operation in channels where applicable) a Servo modeling and comparison to actual electrohydraulic servovalve ( EHSV ) e Control n transmission and...AFWAL-TR-83-3041 A A STUDY OF DIGITALLY CONTROLLED FLIGHT CONTROL ACTUATION HOWARD H. BELMONT HR TEXTRON INC. 25200 WEST RYE CANYON ROAD VALENCIA

  2. Analysis of plasma position control for DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Takase, Haruhiko, E-mail: takase.haruhiko@jaea.go.jp [International Fusion Energy Research Centre, 2-166, Obuchi, Rokkasho, Aomori 0393212 (Japan); Japan Atomic Energy Agency, 2-166, Obuchi, Rokkasho, Aomori 039-3212 (Japan); Utoh, Hiroyasu [International Fusion Energy Research Centre, 2-166, Obuchi, Rokkasho, Aomori 0393212 (Japan); Japan Atomic Energy Agency, 2-166, Obuchi, Rokkasho, Aomori 039-3212 (Japan); Sakamoto, Yoshiteru [Japan Atomic Energy Agency, 2-166, Obuchi, Rokkasho, Aomori 039-3212 (Japan); Mori, Kazuo; Kudo, Tatsuya [International Fusion Energy Research Centre, 2-166, Obuchi, Rokkasho, Aomori 0393212 (Japan); Japan Atomic Energy Agency, 2-166, Obuchi, Rokkasho, Aomori 039-3212 (Japan); Tobita, Kenji [Japan Atomic Energy Agency, 2-166, Obuchi, Rokkasho, Aomori 039-3212 (Japan)

    2016-11-01

    Highlights: • Plasma position control for DEMO reactor has been investigated using numerical code. • Stabilization effect of conductors and active feedback control are evaluated. • Influence on plasma position control by breeding blanket modules is shown. - Abstract: Plasma position control for DEMO reactor has been investigated using numerical simulation, which consists of plasma equilibrium, eddy current and active feedback control analyses. The stabilization effect of in-vessel components, the influence on the magnetic detector and the power of active feedback control coils are evaluated. Especially, the influence of breeding blanket modules on plasma position control is shown in this paper.

  3. Self-operation type power control device for nuclear reactor

    International Nuclear Information System (INIS)

    Kanbe, Mitsuru.

    1993-01-01

    The device of the present invention operates by sensing the temperature change of a reactor core in all of LMFBR type reactors irrespective of the scale of the reactor core power. That is, a region where liquid poison is filled is disposed at the upper portion and a region where sealed gases are filled is disposed at the lower portion of a pipe having both ends thereof being closed. When the pipe is inserted into the reactor core, the inner diameter of the pipe is determined smaller than a predetermined value so that the boundary between the liquid poison and the sealed gases in the pipe is maintained relative to an assumed maximum acceleration. The sealed gas region is disposed at the reactor core region. If the liquid poison is expanded by the elevation of the reactor core exit temperature, it is moved to the lower gas region, to control the reactor power. Since high reliability can be maintained over a long period of time by this method, it is suitable to FBR reactors disposed in such environments that maintenance can not easily be conducted, such as desserts, isolated islands and undeveloped countries. Further, it is also suitable to ultra small sized nuclear reactors disposed at environments that the direction and the magnitude of gravity are different from those on the ground. (I.S.)

  4. Experience of digital control systems in Scandinavian BWRs

    International Nuclear Information System (INIS)

    Rydahl, I.

    1989-01-01

    Since 1984 digital control systems have been in operation in various Scandinavian BWRs. Examples of such digital control systems are: dual microprocessor based system for complete control of radwaste plant, three channel recirculation control system, and three channel feedwater control system. This paper describes Swedish development from one channel through three channel analog control systems to digital systems. The author describes experience of digital control systems during design, testing, commissioning and operation. The main benefits of digital compared with analog technology are discussed. Especially the outstanding facility of using a built-in process simulator for commissioning and tuning. The use of digital technology in nuclear safety system and future plans are dealt with

  5. Aging considerations for PWR [pressurized water reactor] control rod drive mechanisms and reactor internals

    International Nuclear Information System (INIS)

    Ware, A.G.

    1988-01-01

    This paper describes age-related degradation mechanisms affecting life extension of pressurized water reactor control rod drive mechanisms and reactor internals. The major sources of age-related degradation for control rod drive mechanisms are thermal transients such as plant heatups and cooldowns, latchings and unlatchings, long-term aging effects on electrical insulation, and the high temperature corrosive environment. Flow induced loads, the high-temperature corrosive environment, radiation exposure, and high tensile stresses in bolts all contribute to aging related degradation of reactor internals. Another problem has been wear and fretting of instrument guide tubes. The paper also discusses age-related failures that have occurred to date in pressurized water reactors

  6. Material accountancy and control practice at a research reactor facility

    International Nuclear Information System (INIS)

    Bouchard, J.; Maurel, J.J.; Tromeur, Y.

    1982-01-01

    This session surveys the regulations, organization, and accountancy practice that compose the French State System of Accountancy and Control. Practical examples are discussed showing how inventories are verified at a critical assembly facility and at a materials testing reactor

  7. Automatic control system in the reactor peggy

    International Nuclear Information System (INIS)

    Bertrand, J.; Mourchon, R.; Da Costa, D.; Desandre-Navarre, Ch.

    1967-01-01

    The equipment makes it possible for the reactor to attain a given power automatically and for the power to be maintained around this level. The principle of its operation consists in the changing from one power to another, at constant period, by means of a programmer transforming a power-step request into a voltage variation which is linear with time and which represents the logarithm of the required power. The real power is compared continuously with the required power. Stabilization occurs automatically as soon as the difference between the reactor power and the required power diminishes to a few per cent. (authors) [fr

  8. Decision making in the reactor control room

    International Nuclear Information System (INIS)

    Nelson, W.R.

    1982-01-01

    One of the most important roles of the nuclear reactor operator is that of decision maker. This paper discusses a simple model of the decision process used by the reactor operator. Resources that must be available so that he can perform the decision process are presented. Decision aids which have been investigated at EG and G Idaho, Inc., as part of the LOFT Augmented Operator Capability Program are briefly discussed. Some general concepts of computerized decision aiding are developed, and the promises and pitfalls of such decision aids are explored

  9. Neutronics and mass transport in a chemical reactor associated with controlled thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Dang, V.D.; Steinberg, M.; Lazareth, O.W.; Powell, J.R.

    1976-05-01

    The formation of ozone from oxygen and the dissociation carbon dioxide to carbon monoxide and oxygen is studied in a gamma-neutron chemical process blanket associated with a controlled thermonuclear reactor. Materials used for reactor tube wall will affect the efficiency of the energy absorption by the reactants and consequently the yield of reaction products. Three kinds of materials, aluminum, stainless steel and fiber (Al 2 O 3 )-aluminium are investigated for the tube wall material in the study

  10. Brookhaven Reactor Experiment Control Facility, a distributed function computer network

    International Nuclear Information System (INIS)

    Dimmler, D.G.; Greenlaw, N.; Kelley, M.A.; Potter, D.W.; Rankowitz, S.; Stubblefield, F.W.

    1975-11-01

    A computer network for real-time data acquisition, monitoring and control of a series of experiments at the Brookhaven High Flux Beam Reactor has been developed and has been set into routine operation. This reactor experiment control facility presently services nine neutron spectrometers and one x-ray diffractometer. Several additional experiment connections are in progress. The architecture of the facility is based on a distributed function network concept. A statement of implementation and results is presented

  11. Monitoring and control of anaerobic reactors

    DEFF Research Database (Denmark)

    Pind, Peter Frode; Angelidaki, Irini; Ahring, Birgitte Kiær

    2003-01-01

    measurements are reviewed in detail. In the sequel, possible manipulated variables, such as the hydraulic retention time, the organic loading rate, the sludge retention time, temperature, pH and alkalinity are evaluated with respect to the two main reactor types: high-rate and low-rate. Finally, the different...

  12. Reactor core flow rate control system

    International Nuclear Information System (INIS)

    Sakuma, Hitoshi; Tanikawa, Naoshi; Takahashi, Toshiyuki; Miyakawa, Tetsuya.

    1996-01-01

    When an internal pump is started by a variable frequency power source device, if magnetic fields of an AC generator are introduced after the rated speed is reached, neutron flux high scram occurs by abrupt increase of a reactor core flow rate. Then, in the present invention, magnetic fields for the AC generator are introduced at a speed previously set at which the fluctuation range of the reactor core flow rate (neutron flux) by the start up of the internal pump is within an allowable value. Since increase of the speed of the internal pump upon its start up is suppressed to determine the change of the reactor core flow rate within an allowable range, increase of neutron fluxes is suppressed to enable stable start up. Then, since transition boiling of fuels caused by abrupt decrease of the reactor core flow rate upon occurrence of abnormality in an external electric power system is prevented, and the magnetic fields for the AC generator are introduced in such a manner to put the speed increase fluctuation range of the internal pump upon start up within an allowable value, neutron flux high scram is not caused to enable stable start-up. (N.H.)

  13. Adaptive Controller Design for Continuous Stirred Tank Reactor

    OpenAIRE

    K. Prabhu; V. Murali Bhaskaran

    2014-01-01

    Continues Stirred Tank Reactor (CSTR) is an important issue in chemical process and a wide range of research in the area of chemical engineering. Temperature Control of CSTR has been an issue in the chemical control engineering since it has highly non-linear complex equations. This study presents problem of temperature control of CSTR with the adaptive Controller. The Simulation is done in MATLAB and result shows that adaptive controller is an efficient controller for temperature control of C...

  14. Digital Single-Phase Power-Factor Controller

    Science.gov (United States)

    Dabney, R. W.

    1983-01-01

    Digital circuit has faster response to load changes. Digital power-factor controller senses changing motor-load torques by sampling open-circuit voltage across gate-controlled silicon switch. Circuit responds more rapidly to hanging loads than analog power-factor controllers because no low-pass filter is in feedback loop.

  15. Automatic power control for a pressurized water reactor

    International Nuclear Information System (INIS)

    Hah, Yung Joon

    1994-02-01

    During a normal operation of a pressurized water reactor (PWR), the reactivity is controlled by control rods, boron, and the average temperature of the primary coolant. Especially in load follow operation, the reactivity change is induced by changes in power level and effects of xenon concentration. The control of the core power distribution is concerned, mainly, with the axial power distribution which depends on insertion and withdrawal of the control rods resulting in additional reactivity compensation. The utilization of part strength control element assemblies (PSCEAs) is quite appropriate for a control of the power distribution in the case of Yonggwang Nuclear Unit 3 (YGN Unit 3). However, control of the PSCEAs is not automatic, and changes in the boron concentration by dilution/boration are done manually. Thus, manual control of the PSCEAs and the boron concentration require the operator's experience and knowledge for a successful load follow operation. In this thesis, the new concepts have been proposed to adapt for an automatic power control in a PWR. One of the new concepts is the mode K control, another is a fuzzy power control. The system in mode K control implements a heavy-worth bank dedicated to axial shape control, independent of the existing regulating banks. The heavy bank provides a monotonic relationship between its motion and the axial power shape change, which allows automatic control of the axial power distribution. And the mode K enables precise regulation, by using double closed-loop control of the reactor coolant temperature and the axial power difference. Automatic reactor power control permits the nuclear power plant to accommodate the load follow operations, including frequency control, to respond to the grid requirements. The mode K reactor control concepts were tested using simulation responses of a Korean standardized 1000-MWe PWR which is a reference plant for the YGN Unit 3. The simulation results illustrate that the mode K would be

  16. Computer-based regulating control system for the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Johnson, M.R.

    1983-01-01

    This paper describes a new control system which has recently been designed and installed at the Advanced Test Reactor at INEL, replacing an older system that had been in service for some 17 years. Based on modern digital technology, the new system provides improved capability, reliability, and an enhanced man/machine interface that includes comprehensive failure and error messages and voice synthesis. In addition to control functions, and transparent to the operator, the system performs continual on-line checks to sense subsystem failures and takes appropriate automatic action. In the maintenance mode, service technicians can carry on a dialog with the controller to quickly identify faulty components. The operational capabilities of the new system are summarized, and reactor operator training, experience, and acceptance of the system are discussed

  17. Performance of a digital reactivity-meter (009-NC/1-IPEN) in initial test programs for research and power reactor

    International Nuclear Information System (INIS)

    Moreira, J.M.L.; Ferreira, P.S.B.; Pontes, E.W.; Maiorino, J.R.; Soares, A.J.

    1987-01-01

    This paper describes the digital reactivity-meter (009-NC/1-IPEN) built at the IPEN/CNEN-SP for the start-up tests of Angra-I power station. It is also being used in the IEA-R1 research reactor for evaluating control rod worth and the various reactivity coefficients. The equipment is composed of two main parts: an electronic module with 12 bit A/D and D/A interfaces, a picoamperimeter and several microprocessors and the micro-computer in which is solved the inverse kinetics equation to obtain the reactivity as a function of time. The results obtained demonstrate the accuracy and the practicability of the reactivity meter. (Author)

  18. Control rod studies in small and medium sized fast reactors

    International Nuclear Information System (INIS)

    John, T.M.; Mohanakrishnan, P.; Mahalakshmi, B.; Singh, R.S.

    1988-01-01

    Control rods are the primary safety mechanism in the operation of fast reactors. Neutronic parameters associated with the control rods have to be evaluated precisely for studying the behaviour of the reactor under various operating conditions. Control rods are strong neutron absorbers discretely distributed in the reactor core. Accurate estimation of control rod parameters demand, in principle transport theory solutions in exact geometry. But computer codes for such evaluations usually consume exorbitantly large computer time and memory for even a single parameter evaluation. During the design of reactors, evaluation of these parameters will be required for many configurations of control rods. In this paper, the method used at Indira Gandhi Centre for Atomic Research for estimating the parameters associated with control rods is presented. Diffusion theory solutions were used for computations. A scheme using three dimensional geometry represented by triangular meshes and diffusion theory solutions in few energy groups for control rod parameter evaluation is presented. This scheme was employed in estimating the control rod parameters in a 500 Mw(e) fast reactor. Error due to group collapsing is estimated by comparing with 25 group calculations in three dimensions for typical cases. (author). 5 refs, 4 figs, 3 tabs

  19. Modern control technology for improved nuclear reactor performance

    International Nuclear Information System (INIS)

    Oakes, L.C.

    1986-01-01

    One of the main complaints leveled at reactor control systems by utility spokesmen is complexity. One only has to look inside a power reactor control room to appreciate this viewpoint. The high reliability and versatility of modern microprocessors makes possible distributed control systems with only performance data and abnormal conditions being relayed to the control room. In a sense, this emulates the human-body control system where routine repetitive actions are handled in an involuntary manner. The significance of expert systems to the nuclear reactor control and safety systems is their ability to capture human and other expertise and make it available, upon demand, and under almost all circumstances. Thus, human problem-solving skills acquired by the learning process over a long period of time can be captured and employed with the reliability inherent in computers. This is especially important in nuclear plants when human operators are burdened by stress and emotional factors that have a dramatic effect on performance level

  20. Instrumentation and control improvements at Experimental Breeder Reactor II

    International Nuclear Information System (INIS)

    Christensen, L.J.; Planchon, H.P.

    1993-01-01

    The purpose of this paper is to describe instrumentation and control (I ampersand C) system improvements at Experimental Breeder Reactor 11 (EBR-11). The improvements are focused on three objectives; to keep the reactor and balance of plant (BOP) I ampersand C systems at a high level of reliability, to provide diagnostic systems that can provide accurate information needed for analysis of fuel performance, and to provide systems that will be prototypic of I ampersand C systems of the next generation of liquid metal reactor (LMR) plants

  1. Application of fuzzy logic control system for reactor feed-water control

    International Nuclear Information System (INIS)

    Iijima, T.; Nakajima, Y.

    1994-01-01

    The successful actual application of a fuzzy logic control system to the a nuclear Fugen nuclear power reactor is described. Fugen is a heavy-water moderated, light-water cooled reactor. The introduction of fuzzy logic control system has enabled operators to control the steam drum water level more effectively in comparison to a conventional proportional-integral (PI) control system

  2. Tritium containment of controlled thermonuclear fusion reactor

    International Nuclear Information System (INIS)

    Tanaka, Yoshihisa; Tsukumo, Kiyohiko; Suzuki, Tatsushi

    1979-01-01

    It is well known that tritium is used as the fuel for nuclear fusion reactors. The neutrons produced by the nuclear fusion reaction of deuterium and tritium react with lithium in blankets, and tritium is produced. The blankets reproduce the tritium consumed in the D-T reaction. Tritium circulates through the main cooling system and the fuel supply and evacuation system, and is accumulated. Tritium is a radioactive substance emitting β-ray with 12.6 year half-life, and harmful to human bodies. It is an isotope of hydrogen, and apt to diffuse and leak. Especially at high temperature, it permeates through materials, therefore it is important to evaluate the release of tritium into environment, to treat leaked tritium to reduce its release, and to select the method of containing tritium. The permeability of tritium and its solubility in structural materials are discussed. The typical blanket-cooling systems of nuclear fusion reactors are shown, and the tungsten coating of steam generator tubes and tritium recovery system are adopted for reducing tritium leak. In case of the Tokamak type reactor of JAERI, the tritium recovery system is installed, in which the tritium gas produced in blankets is converted to tritium steam with a Pd-Pt catalytic oxidation tower, and it is dehydrated and eliminated with a molecular sieve tower, then purified and recovered. (Kako, I.)

  3. Reactivity control of nuclear power reactors: new options

    International Nuclear Information System (INIS)

    Alcala, F.

    1984-01-01

    Some actual aspects (referring to economy, non-proliferation and environmental impact) of nuclear power reactors has been analyzed from the point of view of the reactivity control physics. Specially studied have been the physical mechanisms related with the spectral shift control method and their general positive effects on those aspects. The analysis carried out suggested the application of the above method of control to reactors with non-hydrogenous fuel cells, which are mainly characterized by their high moderator/fuel ratio. Finally three different types of such fuel cells are presented and some results about one of them (belonging to a PHWR controlled by graphite rods) are given. (author)

  4. A master-follower type distributed scheme for reactor inlet temperature control

    International Nuclear Information System (INIS)

    Garcia, H.E.; Dean, E.M.; Vilim, R.B.

    1995-01-01

    This paper describes the implementation of a computer-based controller for regulating reactor inlet temperature in a pool-type power plant. The elements of the control system are organized in a master-follower hierarchical architecture that takes advantage of existing in-plant hardware and software to minimize the need for plant modifications. Low level control algorithms are executed on existing local digital controllers (followers) with the high level algorithms executed on a new plant supervisory computer (master). A distributed computing strategy provides integration of the existing and additional computer platforms. The control system operates by having the master controller first estimate the secondary sodium flow needed to achieve a given reactor inlet temperature. The estimated flow is then used as a setpoint by the follower controller to regulate sodium flow using a motor-generator pump set. The control system has been implemented in a Hardware-In-the-Loop (FM) setup and qualified for operation in the Experimental Breader reactor 11 of Argonne National Laboratory. Some HIL results are provided

  5. Documenting control system functionality for digital control implementations

    International Nuclear Information System (INIS)

    Harber, J.; Borairi, M.; Tikku, S.; Josefowicz, A.

    2006-01-01

    In past CANDU designs, plant control was accomplished by a combination of digital control computers, analogue controllers, and hardwired relay logic. Functionality for these various control systems, each using different hardware, was documented in varied formats such as text based program specifications, relay logic diagrams, and other various specification documents. The choice of formats was influenced by the hardware used and often required different specialized skills for different applications. The programmable electronic systems in new CANDU designs are realized in a manner consistent with latest international standards (e.g., the IEC 61513 standard). New CANDU designs make extensive use of modern digital control technology, with the benefit that functionality can be implemented on a limited number of control platforms, reducing development and maintenance cost. This approach can take advantage of tools that allow the plant control system functional and performance requirements to be documented using graphical representations. Modern graphical methods supplemented by information databases can be used to provide a clear and comprehensive set of requirements for software and system development. Overview diagrams of system functionality provide a common understanding of the system boundaries and interfaces. Important requirements are readily traced through the development process. This improved reviewability helps to ensure consistency with the safety and and production design requirements of the system. Encapsulation of commonly used functions into custom-defined function blocks, such as typical motor control centre interfaces, process interlocks, median selects etc, eases the burden on designers to understand and analyze the detailed functionality of each instance of use of this logic. A library of encapsulated functions will be established for complex functions that are reused in the control logic development. By encapsulation and standardisation of such

  6. Modelling of Control Bars in Calculations of Boiling Water Reactors

    International Nuclear Information System (INIS)

    Khlaifi, A.; Buiron, L.

    2004-01-01

    The core of a nuclear reactor is generally composed of a neat assemblies of fissile material from where neutrons were descended. In general, the energy of fission is extracted by a fluid serving to cool clusters. A reflector is arranged around the assemblies to reduce escaping of neutrons. This is made outside the reactor core. Different mechanisms of reactivity are generally necessary to control the chain reaction. Manoeuvring of Boiling Water Reactor takes place by controlling insertion of absorbent rods to various places of the core. If no blocked assembly calculations are known and mastered, blocked assembly neutronic calculation are delicate and often treated by case to case in present studies [1]. Answering the question how to model crossbar for the control of a boiling water reactor ? requires the choice of a representation level for every chain of variables, the physical model, and its representing equations, etc. The aim of this study is to select the best applicable parameter serving to calculate blocked assembly of a Boiling Water Reactor. This will be made through a range of representative configurations of these reactors and used absorbing environment, in order to illustrate strategies of modelling in the case of an industrial calculation. (authors)

  7. Optimized Control Rods of the BR2 Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kalcheva, Silva; Koonen, E.

    2007-09-15

    At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

  8. Optimized Control Rods of the BR2 Reactor

    International Nuclear Information System (INIS)

    Kalcheva, Silva; Koonen, E.

    2007-01-01

    At the present time the BR-2 reactor uses control elements with cadmium as neutron absorbing part. The lower section of the control element is a beryllium assembly cooled by light water. Due to the burn up of the lower end of the cadmium section during the reactor operation, the presently used rods for reactivity control of the BR-2 reactor have to be replaced by new ones. Considered are various types Control Rods with full active part of the following materials: cadmium (Cd), hafnium (Hf), europium oxide (Eu2O3) and gadolinium (Gd2O3). Options to decrease the burn up of the control rod material in the hot spot, such as use of stainless steel in the lower active part of the Control Rod are discussed. Comparison with the characteristics of the presently used Control Rods types is performed. The changing of the characteristics of different types Control Rods and the perturbation effects on the reactor neutronics during the BR-2 fuel cycle are investigated. The burn up of the Control Rod absorbing material, total and differential control rods worth, macroscopic and effective microscopic absorption cross sections, fuel and reactivity evolution are evaluated during approximately 30 operating cycles.

  9. Development of web based courseware and digital education platform architecture for nuclear reactor operation experiment

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M. H.; Lee, W. K.; Kim, S. N.; Suh, D. Y. [Kyung Hee Univ., Seoul (Korea)

    2001-01-01

    In this research, two kinds of work were performed for the knowledge base structure of university education environment. The first work is to develop a web-based courseware of 'reactor experiment'. Eight chapter were designed - guide and introduction, system tour on nuclear power plant, reactor kinetics theory, application of reactor kinetics, criticality measurement experiment, reactivity measurement, load-following operation transient, reactor operation experiment. At this point, 5 chapters were completed and 3 chapters are under construction with expectation of near-term completion. A html courseware files were operated on PC LINUX and on-line login can be done on the site 'http://cylex.kyunghee.ac.kr'. Each chapter consist of lecture note, lecture slides, self-diagnostic quiz, pass/fail exam and cyber simulator. The second task of this project was to build a cyber lecture and experiment space(CyLEX) to operate a developed courseware on it. 30 seats classroom was completed in the engineering bldg at Kyung Hee University. A advanced space - CyLEX was equipped with computers, electronic board, beam projector, digital camera OHP, etc. Courseware was loaded on the server and opened to public on 24 hour base. Another function of classroom is a capability of lecturer to monitor and command of student computers. Development on system was focused on operation for on-site lecture with intra-net. However, remote on-line class for inter-net can be open to the public under the limitation of transmission speed via internet gateways. 40 figs., 1 tabs. (Author)

  10. The development of digital oscilloscope control software in nuclear measurement

    International Nuclear Information System (INIS)

    Pu Minghui; Tian Geng; Li Xianyou

    2004-01-01

    This essay presents the development of an all-purpose digital oscilloscope control software on Windows 95/98 OS. The background and method are discussed in detail, together with the function and characteristics of the software. With the use of this software, a single PC can control several digital oscilloscopes. Solution of main problems encountered in the development is also discussed. (authors)

  11. MATLAB simulation for an experimental setup of digital feedback control

    International Nuclear Information System (INIS)

    Zheng Lifang; Liu Songqiang

    2005-01-01

    This paper describes the digital feedback simulation using MATLAB for an experimental accelerator control setup. By analyzing the plant characteristic in time-domain and frequency-domain, a guideline for design of digital filter and PID controller is derived. (authors)

  12. Regulatory perspective on digital instrumentation and control systems for future advanced nuclear power plants

    International Nuclear Information System (INIS)

    Chiramal, M.

    1993-01-01

    This paper deals with the question of using digital technology in instrumentation and control systems for modern nuclear power reactors. The general opinion in the industry and among NRC staff is that such technology provides the opportunity for enhanced safety and reliable reactor operations. The major concern is the safe application of this technology so as to avoid common mode or common cause failures in systems. There are great differences between digital and analog system components. SECY-91-292 identifies some general regulatory concerns with regard to digital systems. There is clearly a lack of adequate regulatory direction on the application of digital equipment at this time, but the issue is being addressed by the industry, outside experts, and NRC staff. NRC staff presents a position on the issue of defense-in-depth and diversity with regard to insuring plant safety. Independent manual controls and readouts must be available to allow safe shutdown and monitoring of the plant in the event of safety system failures

  13. Human factors evaluation of the engineering test reactor control room

    International Nuclear Information System (INIS)

    Banks, W.W.; Boone, M.P.

    1981-03-01

    The Reactor and Process Control Rooms at the Engineering Test Reactor were evaluated by a team of human factors engineers using available human factors design criteria. During the evaluation, ETR, equipment and facilities were compared with MIL-STD-1472-B, Human Engineering design Criteria for Military Systems. The focus of recommendations centered on: (a) displays and controls; placing displays and controls in functional groups; (b) establishing a consistent color coding (in compliance with a standard if possible); (c) systematizing annunciator alarms and reducing their number; (d) organizing equipment in functional groups; and (e) modifying labeling and lines of demarcation

  14. Use of university research reactors to teach control engineering

    International Nuclear Information System (INIS)

    Bernard, J.A.

    1991-01-01

    University research reactors (URRs) have provided generations of students with the opportunity to receive instruction and do hands-on work in reactor dynamics, neutron scattering, health physics, and neutron activation analysis. Given that many URRs are currently converting to programmable control systems, the opportunity now exists to provide a similar learning experience to those studying systems control engineering. That possibility is examined here with emphasis on the need for the inclusion of experiment in control engineering curricula, the type of activities that could be performed, and safety considerations

  15. Automated power control system for reactor TRIGA PUSPATI

    Science.gov (United States)

    Ghazali, Anith Khairunnisa; Minhat, Mohd Sabri; Hassan, Mohd Khair

    2017-01-01

    Reactor TRIGA PUSPATI (RTP) Mark II type undergoes safe operation for more than 30 years and the only research reactor exists in Malaysia. The main safety feature of Instrumentation and Control (I&C) system design is such that any failure in the electronic, or its associated components, does not lead to an uncontrolled rate of reactivity. The existed controller using feedback approach to control the reactor power. This paper introduces proposed controllers such as Model Reference Adaptive Control (MRAC) and Proportional Integral Derivatives (PID) controller for the RTP simulation. In RTP, the most important considered parameter is the reactor power and act as nervous system. To design a controller for complex plant like RTP is quite difficult due to high cost and safety factors cause by the failure of the controller. Furthermore, to overcome these problems, a simulator can be used to replace functions the hardware and test could then be simulated using this simulator. In order to find the best controller, several controllers were proposed and the result will be analysed for study the performances of the controller. The output result will be used to find out the best RTP power controller using MATLAB/Simulink and gives result as close as the real RTP performances. Currently, the structures of RTP was design using MATLAB/Simulink tool that consist of fission chamber, controller, control rod position, height-to-worth of control rods and a RTP model. The controller will control the control rod position to make sure that the reactivity still under the limitation parameter. The results given from each controller will be analysed and validated through experiment data collected from RTP.

  16. Inherent reactor power controller for a metal-fueled ALMR

    International Nuclear Information System (INIS)

    Wood, R.T.; Wilson, T.L. Jr.

    1990-01-01

    Inherent power control for metal-fueled ALMR designs involves using reactivity thermal feedback effects to control reactor power. This paper describes how, using classical control design techniques, a control system for normal load following maneuvers was deigned for a pool-type ALMR. This design provides active control of power removal in the balance of plant, direct control of selected primary and intermediate loop temperatures, and passive control of reactor power. The inherent stability of the strong, fast reactivity feedback effects bring heat production in the core into balance with the heat removal system temperatures, which are controlled to meet power demand. A simulation of the control system successfully responded to a 10% step change in power demand by changing power at an acceptable rate without causing large temperature fluctuations or exceeding thermal limits

  17. Multilayer robust control for safety enhancement of reactor operations

    International Nuclear Information System (INIS)

    Edwards, R.M.; Lee, K.Y.; Ray, A.

    1991-01-01

    A novel concept of reactor power and temperature control has been recently reported in which a conventional output feedback controller is embedded within a state feedback setting. The embedded output feedback controller at the inner layer largely compensates for plant modeling uncertainties and external disturbances, and the outer layer generates an optimal control signal via feedback of the estimated plant states. A major advantage of this embedded architecture is the robustness of the control system relative to parametric and nonparametric uncertainties and thus the opportunity for designing fault-accommodating control algorithms to improve reactor operations and plant safety. The paper illustrates the architecture of the state-feedback-assisted classical (SFAC) control, which utilizes an embedded output feedback controller designed via classical techniques. It demonstrates the difference between the performance of conventional state feedback control and SFAC by examining the sensitivity of the dominant eigenvalues of the individual closed-loop systems

  18. Vacuum pumping for controlled thermonuclear reactors

    International Nuclear Information System (INIS)

    Watson, J.S.; Fisher, P.W.

    1976-01-01

    Thermonuclear reactors impose unique vacuum pumping problems involving very high pumping speeds, handling of hazardous materials (tritium), extreme cleanliness requirements, and quantitative recovery of pumped materials. Two principal pumping systems are required for a fusion reactor, a main vacuum system for evacuating the torus and a vacuum system for removing unaccelerated deuterium from neutral beam injectors. The first system must pump hydrogen isotopes and helium while the neutral beam system can operate by pumping only hydrogen isotopes (perhaps only deuterium). The most promising pumping techniques for both systems appear to be cryopumps, but different cryopumping techniques can be considered for each system. The main vacuum system will have to include cryosorption pumps cooled to 4.2 0 K to pump helium, but the unburned deuterium-tritium and other impurities could be pumped with cryocondensation panels (4.2 0 K) or cryosorption panels at higher temperatures. Since pumping speeds will be limited by conductance through the ducts and thermal shields, the pumping performance for both systems will be similar, and other factors such as refrigeration costs are likely to determine the choice. The vacuum pumping system for neutral beam injectors probably will not need to pump helium, and either condensation or higher temperature sorption pumps can be used

  19. Emergency facility control device for nuclear reactor

    International Nuclear Information System (INIS)

    Ikehara, Morihiko.

    1981-01-01

    Purpose: To increase the reliability of a nuclear reactor by allowing an emergency facility to be manually started and stopped to make its operation more convenient and eliminate the possibility of erroneous operation in an emergency. Constitution: There are provided a first water level detector for detecting a level lower than the first low water level in a reactor container and a second water level detector for detecting a level lower than the second low water level lower than the first low water level, and an emergency facility can be started and stopped manually only when the level is higher than the second low water level, but the facility will be started regardless of the state of the manual operation when the level is lower than the second low water level. Thus, the emergency facility can be started by manual operation, but will be automatically started so as to secure the necessary minimum operation if the level becomes lower than the second low water level and the stopping operation thereafter is forgotten. (Kamimura, M.)

  20. On some control problems of dynamic of reactor

    Science.gov (United States)

    Baskakov, A. V.; Volkov, N. P.

    2017-12-01

    The paper analyzes controllability of the transient processes in some problems of nuclear reactor dynamics. In this case, the mathematical model of nuclear reactor dynamics is described by a system of integro-differential equations consisting of the non-stationary anisotropic multi-velocity kinetic equation of neutron transport and the balance equation of delayed neutrons. The paper defines the formulation of the linear problem on control of transient processes in nuclear reactors with application of spatially distributed actions on internal neutron sources, and the formulation of the nonlinear problems on control of transient processes with application of spatially distributed actions on the neutron absorption coefficient and the neutron scattering indicatrix. The required control actions depend on the spatial and velocity coordinates. The theorems on existence and uniqueness of these control actions are proved in the paper. To do this, the control problems mentioned above are reduced to equivalent systems of integral equations. Existence and uniqueness of the solution for this system of integral equations is proved by the method of successive approximations, which makes it possible to construct an iterative scheme for numerical analyses of transient processes in a given nuclear reactor with application of the developed mathematical model. Sufficient conditions for controllability of transient processes are also obtained. In conclusion, a connection is made between the control problems and the observation problems, which, by to the given information, allow us to reconstruct either the function of internal neutron sources, or the neutron absorption coefficient, or the neutron scattering indicatrix....

  1. Plant Control of the High Performance Light Water Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schlagenhaufer, Marc; Starflinger, J.; Schulenberg, T. [Institute for Nuclear and Energy Technologies, Forschungszentrum Karlsruhe GmbH, Hermann-von-Helmholtz-Platz 1, Eggenstein-Leopoldshafen, Baden-Wuertemberg 76344 (Germany)

    2009-06-15

    The latest design concept of the High Performance Light Water Reactor (HPLWR) includes a thermal core in which supercritical water at 25 MPa inlet pressure is heated up from 280 deg. C reactor inlet temperature to 500 deg. C core exit temperature in three steps with intermediate coolant mixing to minimize peak cladding temperatures of the fuel rods. A direct supercritical steam cycle of the HPLWR has been designed with high, intermediate and low pressure turbines with a single reheat to 441 deg. C at 4.04 MPa pressure. Three low pressure pre-heaters and four high pressure pre-heaters are foreseen to achieve the envisaged reactor inlet temperature of 280 deg. C at full load. A feedwater tank of 603 m{sup 3} at 0.55 MPa pressure serves as an accumulator for normal and accidental conditions. The steam cycle has been modelled with APROS, developed by VTT Finland, to provide thermodynamic data and cycle efficiency values under full load and part load operation conditions as well as the transient response to load changes. A plant control system has been designed in which the reactor inlet pressure is controlled by the turbine valve, the reactor power is controlled by the feedwater pumps while the life steam temperature is controlled by control rods, and the reheat temperature is controlled by the reheater valve. Neglecting the reactivity control, the core power can also be treated as input parameter such that the life steam temperature is directly controlled by the feedwater mass flow. The plant control can handle all loading and de-loading cycles including complete shut down. A constant pressure at reactor inlet is foreseen for all load cases. Peak temperatures of the fuel pins are checked with a simplified core model. Two shut down procedures starting at 50% load are presented. A reactor scram with turbine states the safe shut down of the whole plant. To avoid hard material temperature changes, a controlled shut down procedure is designed. The rotational speed of the

  2. TREAT [Transient Reactor Test Facility] reactor control rod scram system simulations and testing

    International Nuclear Information System (INIS)

    Solbrig, C.W.; Stevens, W.W.

    1990-01-01

    Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent ampersand Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs

  3. RAPID-L Highly Automated Fast Reactor Concept Without Any Control Rods (1) Reactor concept and plant dynamics analyses

    International Nuclear Information System (INIS)

    Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Iwamura, Takamichi

    2002-01-01

    The 200 kWe uranium-nitride fueled lithium cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for Lunar base power system. It is one of the variants of RAPID (Refueling by All Pins Integrated Design), fast reactor concept, which enable quick and simplified refueling. The essential feature of RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small size reactor core, 2700 fuel pins are integrated altogether and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 years. Unique challenges in reactivity control systems design have been attempted in RAPID-L concept. The reactor has no control rod, but involves the following innovative reactivity control systems: Lithium Expansion Modules (LEM) for inherent reactivity feedback, Lithium Injection Modules (LIM) for inherent ultimate shutdown, and Lithium Release Modules (LRM) for automated reactor startup. All these systems adopt lithium-6 as a liquid poison instead of B 4 C rods. In combination with LEMs, LIMs and LRMs, RAPID-L can be operated without operator. This is the first reactor concept ever established in the world. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, RAPID-L reactor concept and its transient characteristics are presented. (authors)

  4. Atmosphere control device in a reactor container

    International Nuclear Information System (INIS)

    Kasagi, Toru.

    1976-01-01

    Object: To permit maintenance and inspection of the apparatus to be carried out readily and safely without possibility of exposure to radioactive rays by providing a heat exchanger integrally on the outside of the container by means of air paths. Structure: Support structure, shield and the like are disposed to concentrically surround the reactor and are accommodated in the container. The outer side of the container is provided with an integral casing, within which a heat exchanger is disposed. The container wall covered by the casing is provided with forward and backward air paths, which are connected to the heat exchanger by respective ducts. Ducts are led to the outside of the casing, and a fan and filter are provided in the ducts. (seki, T.)

  5. Pressurised Water Reactor Control by the Hierarchical Method

    Directory of Open Access Journals (Sweden)

    Ilkka Leikkonen

    1987-04-01

    Full Text Available A simple version of the hierarchical optimization method is used to solve the control problem for the power distribution of a pressurized water reactor. The control period is about twenty hours. The control objectives include the total power, power distribution and use of boron. The controllers are a rod bank, soluble boron in the coolant and the coolant temperature deviation. A one-dimensional non-linear core model is used, with full xenon-iodine dynamics.

  6. An Embedded Based Digital Controller for Thermal Process

    Directory of Open Access Journals (Sweden)

    A. Lakshmi Sangeetha

    2008-01-01

    Full Text Available This paper describes a low cost virtual instrumentation (VI system to monitor and control the electrically heated water bath temperature. The PIC16F877 based digital microcontroller is used as thermostat which controls and monitors the temperature. The digital controller also allows the user to modify the sensor (PT100 calibration data values if necessary. The developed programmable on/off control function provides on-line display of measuring temperature, set point as well as the control function output plots through the parallel port. This bus interaction is realized in Visual Basic/Assembly Language and uses a 16 bit, 10 ms sampling analog-to-digital converter (ADS 7805 for monitoring and controlling the parameters of the temperature local digital controller.

  7. Light Water Reactor Sustainability Program Advanced Instrumentation, Information, and Control Systems Technologies Technical Program Plan for 2013

    Energy Technology Data Exchange (ETDEWEB)

    Hallbert, Bruce [Idaho National Laboratory (INL), Idaho Falls, ID (United States); Thomas, Ken [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2014-09-01

    Reliable instrumentation, information, and control (II&C) systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration.

  8. Light Water Reactor Sustainability Program Advanced Instrumentation, Information, and Control Systems Technologies Technical Program Plan for FY 2016

    Energy Technology Data Exchange (ETDEWEB)

    Hallbert, Bruce Perry [Idaho National Lab. (INL), Idaho Falls, ID (United States); Thomas, Kenneth David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-10-01

    Reliable instrumentation, information, and control (II&C) systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration.

  9. Coordinate control of integral reactor based on single neuron PID controller

    International Nuclear Information System (INIS)

    Liu Yan; Xia Hong

    2014-01-01

    As one of the main type of reactors in the future, the development of the integral reactor has attracted worldwide attention. On the basis of understanding the background of the integral reactor, the author will be familiar with and master the power control of reactor and the feedwater flow control of steam generator, and the speed control of turbine (turbine speed control is associated with the turbine load control). According to the expectative program 'reactor power following turbine load' of the reactor, it will make coordinate control of the three and come to a overall control scheme. The author will use the supervisory learning algorithm of Hebb for single neuron PID controller with self-adaptation to study the coordinate control of integral reactor. Compared with conventional PI or PID controller, to a certain extent, it solves the problems that traditional PID controller is not easy to tune real-time parameters and lack of effective control for a number of complex processes and slow-varying parameter systems. It improves the security, reliability, stability and flexibility of control process and achieves effective control of the system. (authors)

  10. Cyber security for remote monitoring and control of small reactors

    International Nuclear Information System (INIS)

    Trask, D.; Jung, C.; MacDonald, M.

    2014-01-01

    There is growing international interest and activity in the development of small nuclear reactor technology with a number of vendors interested in building small reactors in Canada to serve remote locations. A common theme of small reactor designs proposed for remote Canadian locations is the concept of a centrally located main control centre operating several remotely located reactors via satellite communications. This theme was echoed at a recent IAEA conference where a recommendation was made to study I&C for remotely controlled small modular reactors, including satellite links and cyber security. This paper summarizes the results of an AECL-CNSC research project to analyze satellite communication technologies used for remote monitoring and control functions in order to provide cyber security regulatory considerations. The scope of this research included a basic survey of existing satellite communications technology and its use in industrial control applications, a brief history of satellite vulnerabilities and a broad review of over 50 standards, guidelines, and regulations from recognized institutions covering safety, cyber security, and industrial communication networks including wireless communications in general. This paper concludes that satellite communications should not be arbitrarily excluded by standards or regulation from use for the remote control and monitoring of small nuclear reactors. Instead, reliance should be placed on processes that are independent of any particular technology, such as reducing risks by applying control measures and demonstrating required reliability through good design practices and testing. Ultimately, it is compliance to well-developed standards that yields the evidence to conclude whether a particular application that uses satellite communications is safe and secure. (author)

  11. REVIEW OF NRC APPROVED DIGITAL CONTROL SYSTEMS ANALYSIS

    International Nuclear Information System (INIS)

    Markman, D.W.

    1999-01-01

    Preliminary design concepts for the proposed Subsurface Repository at Yucca Mountain indicate extensive reliance on modern, computer-based, digital control technologies. The purpose of this analysis is to investigate the degree to which the U. S. Nuclear Regulatory Commission (NRC) has accepted and approved the use of digital control technology for safety-related applications within the nuclear power industry. This analysis reviews cases of existing digitally-based control systems that have been approved by the NRC. These cases can serve as precedence for using similar types of digitally-based control technologies within the Subsurface Repository. While it is anticipated that the Yucca Mountain Project (YMP) will not contain control systems as complex as those required for a nuclear power plant, the review of these existing NRC approved applications will provide the YMP with valuable insight into the NRCs review process and design expectations for safety-related digital control systems. According to the YMP Compliance Program Guidance, portions of various NUREGS, Regulatory Guidelines, and nuclear IEEE standards the nuclear power plant safety related concept would be applied to some of the designs on a case-by-case basis. This analysis will consider key design methods, capabilities, successes, and important limitations or problems of selected control systems that have been approved for use in the Nuclear Power industry. An additional purpose of this analysis is to provide background information in support of further development of design criteria for the YMP. The scope and primary objectives of this analysis are to: (1) Identify and research the extent and precedence of digital control and remotely operated systems approved by the NRC for the nuclear power industry. Help provide a basis for using and relying on digital technologies for nuclear related safety critical applications. (2) Identify the basic control architecture and methods of key digital control

  12. Reactor power automatically controlling method and device for BWR type reactor

    International Nuclear Information System (INIS)

    Murata, Akira; Miyamoto, Yoshiyuki; Tanigawa, Naoshi.

    1997-01-01

    For an automatic control for a reactor power, when a deviation exceeds a predetermined value, the aimed value is kept at a predetermined value, and when the deviation is decreased to less than the predetermined value, the aimed value is increased from the predetermined value again. Alternatively, when a reactor power variation coefficient is decreased to less than a predetermine value, an aimed value is maintained at a predetermined value, and when the variation coefficient exceeds the predetermined value, the aimed value is increased. When the reactor power variation coefficient exceeds a first determined value, an aimed value is increased to a predetermined variation coefficient, and when the variation coefficient is decreased to less than the first determined value and also when the deviation between the aimed value and an actual reactor power exceeds a second determined value, the aimed value is maintained at a constant value. When the deviation is increased or when the reactor power variation coefficient is decreased, since the aimed value is maintained at predetermined value without increasing the aimed value, the deviation is not increased excessively thereby enabling to avoid excessive overshoot. (N.H.)

  13. Instrumentation for nuclear reactor control and protection in France

    International Nuclear Information System (INIS)

    Weill, J.; Remus, L.

    1983-10-01

    The instrumentation for nuclear reactor control and protection is completely made by the French industry. The research and development works are often realized by CEA in the frame of cooperation with EdF (Electricite de France), and societies such as FRAMATOME and NOVATOME and the manufacturers of electronic equipments. In this paper, the main components used in the nuclear instrumentation are described: radiation detectors and electronic equipments for signal processing. There, the control and protection systems manufactured by MERLIN-GERIN for the 900 MWe and 1300 MWe nuclear reactors are described [fr

  14. Model predictive control of a solar-thermal reactor

    Science.gov (United States)

    Saade Saade, Maria Elizabeth

    Solar-thermal reactors represent a promising alternative to fossil fuels because they can harvest solar energy and transform it into storable and transportable fuels. The operation of solar-thermal reactors is restricted by the available sunlight and its inherently transient behavior, which affects the performance of the reactors and limits their efficiency. Before solar-thermal reactors can become commercially viable, they need to be able to maintain a continuous high-performance operation, even in the presence of passing clouds. A well-designed control system can preserve product quality and maintain stable product compositions, resulting in a more efficient and cost-effective operation, which can ultimately lead to scale-up and commercialization of solar thermochemical technologies. In this work, we propose a model predictive control (MPC) system for a solar-thermal reactor for the steam-gasification of biomass. The proposed controller aims at rejecting the disturbances in solar irradiation caused by the presence of clouds. A first-principles dynamic model of the process was developed. The model was used to study the dynamic responses of the process variables and to identify a linear time-invariant model used in the MPC algorithm. To provide an estimation of the disturbances for the control algorithm, a one-minute-ahead direct normal irradiance (DNI) predictor was developed. The proposed predictor utilizes information obtained through the analysis of sky images, in combination with current atmospheric measurements, to produce the DNI forecast. In the end, a robust controller was designed capable of rejecting disturbances within the operating region. Extensive simulation experiments showed that the controller outperforms a finely-tuned multi-loop feedback control strategy. The results obtained suggest that our controller is suitable for practical implementation.

  15. Nuclear reactor remote disconnect control rod coupling indicator

    International Nuclear Information System (INIS)

    Vuckovich, M.

    1977-01-01

    A coupling indicator for use with nuclear reactor control rod assemblies which have remotely disengageable couplings between the control rod and the control rod drive shaft is described. The coupling indicator indicates whether the control rod and the control rod drive shaft are engaged or disengaged. A resistive network, utilizing magnetic reed switches, senses the position of the control rod drive mechanism lead screw and the control rod position indicating tube, and the relative position of these two elements with respect to each other is compared to determine whether the coupling is engaged or disengaged

  16. Advanced control of propylene polimerizations in slurry reactors

    Directory of Open Access Journals (Sweden)

    Bolsoni A.

    2000-01-01

    Full Text Available The objective of this work is to develop a strategy of nonlinear model predictive control for industrial slurry reactors of propylene polymerizations. The controlled variables are the melt index (polymer quality and the amount of unreacted monomer (productivity. The model used in the controller presents a linear dynamics and a nonlinear static gain given by a neuronal network MLP (multilayer perceptron. The simulated performance of the controller was evaluated for a typical propylene polymerization process. It is shown that the performance of the proposed control strategy is much better than the one obtained with the use of linear predictive controllers for setpoint tracking control problems.

  17. Shielding device for control rod in nuclear reactor

    International Nuclear Information System (INIS)

    Sakamaki, Kazuo; Tomatsu, Tsutomu.

    1995-01-01

    The device of the present invention shields radiation emitted from control rods to greatly reduce an operator's radiation exposure even if reactor water level is lowered and the upper portion of the control rod is exposed upon inspection of a BWR type reactor. Namely, a shield assembly has a structure comprising a set of four columnar shields in a two-row and two-column arrangement, which can be inserted into a control rod guide tube. Upon conducting inspection, the control rod is lowered into the control rod guide tube, and in this state, the columnar shields of the shield assembly are inserted to the control rod in the control rod guide tube. With such procedures, the upper portion of the control rod protruded from the control rod guide tube is covered with the shield assembly. As a result, radiation leaked from the control rod is shielded. Accordingly, irradiation in the reactor due to leaked radiation can be prevented thereby enabling to reduce an operator's radiation exposure. (I.S.)

  18. Robust digital controllers for uncertain chaotic systems: A digital redesign approach

    Energy Technology Data Exchange (ETDEWEB)

    Ababneh, Mohammad [Department of Controls, FMC Kongsberg Subsea, FMC Energy Systems, Houston, TX 77067 (United States); Barajas-Ramirez, Juan-Gonzalo [CICESE, Depto. De Electronica y Telecomunicaciones, Ensenada, BC, 22860 (Mexico); Chen Guanrong [Centre for Chaos Control and Synchronization, Department of Electronic Engineering, City University of Hong Kong (China); Shieh, Leang S. [Department of Electrical and Computer Engineering, University of Houston, Houston, TX 77204-4005 (United States)

    2007-03-15

    In this paper, a new and systematic method for designing robust digital controllers for uncertain nonlinear systems with structured uncertainties is presented. In the proposed method, a controller is designed in terms of the optimal linear model representation of the nominal system around each operating point of the trajectory, while the uncertainties are decomposed such that the uncertain nonlinear system can be rewritten as a set of local linear models with disturbed inputs. Applying conventional robust control techniques, continuous-time robust controllers are first designed to eliminate the effects of the uncertainties on the underlying system. Then, a robust digital controller is obtained as the result of a digital redesign of the designed continuous-time robust controller using the state-matching technique. The effectiveness of the proposed controller design method is illustrated through some numerical examples on complex nonlinear systems--chaotic systems.

  19. Digital signal processing in power system protection and control

    CERN Document Server

    Rebizant, Waldemar; Wiszniewski, Andrzej

    2011-01-01

    Digital Signal Processing in Power System Protection and Control bridges the gap between the theory of protection and control and the practical applications of protection equipment. Understanding how protection functions is crucial not only for equipment developers and manufacturers, but also for their users who need to install, set and operate the protection devices in an appropriate manner. After introductory chapters related to protection technology and functions, Digital Signal Processing in Power System Protection and Control presents the digital algorithms for signal filtering, followed

  20. Fuzzy model-based control of a nuclear reactor

    International Nuclear Information System (INIS)

    Van Den Durpel, L.; Ruan, D.

    1994-01-01

    The fuzzy model-based control of a nuclear power reactor is an emerging research topic world-wide. SCK-CEN is dealing with this research in a preliminary stage, including two aspects, namely fuzzy control and fuzzy modelling. The aim is to combine both methodologies in contrast to conventional model-based PID control techniques, and to state advantages of including fuzzy parameters as safety and operator feedback. This paper summarizes the general scheme of this new research project

  1. Implementation of Adaptive Digital Controllers on Programmable Logic Devices

    Science.gov (United States)

    Gwaltney, David A.; King, Kenneth D.; Smith, Keary J.; Monenegro, Justino (Technical Monitor)

    2002-01-01

    Much has been made of the capabilities of FPGA's (Field Programmable Gate Arrays) in the hardware implementation of fast digital signal processing. Such capability also makes an FPGA a suitable platform for the digital implementation of closed loop controllers. Other researchers have implemented a variety of closed-loop digital controllers on FPGA's. Some of these controllers include the widely used proportional-integral-derivative (PID) controller, state space controllers, neural network and fuzzy logic based controllers. There are myriad advantages to utilizing an FPGA for discrete-time control functions which include the capability for reconfiguration when SRAM-based FPGA's are employed, fast parallel implementation of multiple control loops and implementations that can meet space level radiation tolerance requirements in a compact form-factor. Generally, a software implementation on a DSP (Digital Signal Processor) or microcontroller is used to implement digital controllers. At Marshall Space Flight Center, the Control Electronics Group has been studying adaptive discrete-time control of motor driven actuator systems using digital signal processor (DSP) devices. While small form factor, commercial DSP devices are now available with event capture, data conversion, pulse width modulated (PWM) outputs and communication peripherals, these devices are not currently available in designs and packages which meet space level radiation requirements. In general, very few DSP devices are produced that are designed to meet any level of radiation tolerance or hardness. The goal of this effort is to create a fully digital, flight ready controller design that utilizes an FPGA for implementation of signal conditioning for control feedback signals, generation of commands to the controlled system, and hardware insertion of adaptive control algorithm approaches. An alternative is required for compact implementation of such functionality to withstand the harsh environment

  2. Formulation and experimental evaluation of closed-form control laws for the rapid maneuvering of reactor neutronic power

    International Nuclear Information System (INIS)

    Bernard, J.A.

    1989-09-01

    This report describes both the theoretical development and the experimental evaluation of a novel, robust methodology for the time-optimal adjustment of a reactor's neutronic power under conditions of closed-loop digital control. Central to the approach are the 'MIT-SNL Period-Generated Minimum Time Control Laws' which determine the rate at which reactivity should be changed in order to cause a reactor's neutronic power to conform to a specified trajectory. Using these laws, reactor power can be safely raised by five to seven orders of magnitude in a few seconds. The MIT-SNL laws were developed to facilitate rapid increases of neutronic power on spacecraft reactors operating in an SDI environment. However, these laws are generic and have other applications including the rapid recovery of research and test reactors subsequent to an unanticipated shutdown, power increases following the achievement of criticality on commercial reactors, power adjustments on commercial reactors so as to minimize thermal stress, and automated startups. The work reported here was performed by the Massachusetts Institute of Technology under contract to the Sandia National Laboratories. Support was also provided by the US Department of Energy's Division of University and Industry Programs. The work described in this report is significant in that a novel solution to the problem of time-optimal control of neutronic power was identified, in that a rigorous description of a reactor's dynamics was derived in that the rate of change of reactivity was recognized as the proper control signal, and in that extensive experimental trials were conducted of these newly developed concepts on actual nuclear reactors. 43 refs., 118 figs., 11 tabs

  3. Formulation and experimental evaluation of closed-form control laws for the rapid maneuvering of reactor neutronic power

    Energy Technology Data Exchange (ETDEWEB)

    Bernard, J.A. (Massachusetts Inst. of Tech., Cambridge, MA (USA). Nuclear Reactor Lab.)

    1989-09-01

    This report describes both the theoretical development and the experimental evaluation of a novel, robust methodology for the time-optimal adjustment of a reactor's neutronic power under conditions of closed-loop digital control. Central to the approach are the MIT-SNL Period-Generated Minimum Time Control Laws' which determine the rate at which reactivity should be changed in order to cause a reactor's neutronic power to conform to a specified trajectory. Using these laws, reactor power can be safely raised by five to seven orders of magnitude in a few seconds. The MIT-SNL laws were developed to facilitate rapid increases of neutronic power on spacecraft reactors operating in an SDI environment. However, these laws are generic and have other applications including the rapid recovery of research and test reactors subsequent to an unanticipated shutdown, power increases following the achievement of criticality on commercial reactors, power adjustments on commercial reactors so as to minimize thermal stress, and automated startups. The work reported here was performed by the Massachusetts Institute of Technology under contract to the Sandia National Laboratories. Support was also provided by the US Department of Energy's Division of University and Industry Programs. The work described in this report is significant in that a novel solution to the problem of time-optimal control of neutronic power was identified, in that a rigorous description of a reactor's dynamics was derived in that the rate of change of reactivity was recognized as the proper control signal, and in that extensive experimental trials were conducted of these newly developed concepts on actual nuclear reactors. 43 refs., 118 figs., 11 tabs.

  4. Power control of water reactors using nitrogen 16 activity measurements

    International Nuclear Information System (INIS)

    Gariod, R.; Merchie, F.; O'byrne, G.

    1964-01-01

    At the Grenoble Nuclear Research Centre, the open-core swimming pool reactors Melusine (2 MW) and Siloe (15 MW) are controlled at a constant overall power using nitrogen-16 channels. The conventional linear control channels react instantaneously to the rapid power fluctuations, this being necessary for the safety of the reactors, but their power indications are erroneous since they are affected by local deformations of the thermal flux caused by the compensation movements of the control rods. The nitrogen-16 channels on the other hand give an indication of the overall power proportional to the mean fission flux and independent of the rod movements, but their response time is 15 seconds, A constant overall power control is thus possible by a slow correction of the reference signal given by the automatic control governed by thu linear channels by means of a correction term given by the 'N-16' channels: This is done automatically in Melusine and manually in Siloe. (authors) [fr

  5. Control of radioactive material transport in sodium-cooled reactors

    International Nuclear Information System (INIS)

    Brehm, W.F.

    1980-03-01

    The Radioactivity Control Technology (RCT) program was established by the Department of Energy to develop and demonstrate methods to control radionuclide transport to ex-core regions of sodium-cooled reactors. This radioactive material is contained within the reactor heat transport system with any release to the environment well below limits established by regulations. However, maintenance, repair, decontamination, and disposal operations potentially expose plant workers to radiation fields arising from radionuclides transported to primary system components. This paper deals with radioactive material generated and transported during steady-state operation, which remains after 24 Na decay. Potential release of radioactivity during postulated accident conditions is not discussed. The control methods for radionuclide transport, with emphasis on new information obtained since the last Environmental Control Symposium, are described. Development of control methods is an achievable goal

  6. Automatic coolant flow control device for a nuclear reactor assembly

    Science.gov (United States)

    Hutter, E.

    1984-01-27

    A device which controls coolant flow through a nuclear reactor assembly comprises a baffle means at the exit end of said assembly having a plurality of orifices, and a bimetallic member in operative relation to the baffle means such that at increased temperatures said bimetallic member deforms to unblock some of said orifices and allow increased coolant flow therethrough.

  7. Optimization and control of a continuous polymerization reactor

    Directory of Open Access Journals (Sweden)

    L. A. Alvarez

    2012-12-01

    Full Text Available This work studies the optimization and control of a styrene polymerization reactor. The proposed strategy deals with the case where, because of market conditions and equipment deterioration, the optimal operating point of the continuous reactor is modified significantly along the operation time and the control system has to search for this optimum point, besides keeping the reactor system stable at any possible point. The approach considered here consists of three layers: the Real Time Optimization (RTO, the Model Predictive Control (MPC and a Target Calculation (TC that coordinates the communication between the two other layers and guarantees the stability of the whole structure. The proposed algorithm is simulated with the phenomenological model of a styrene polymerization reactor, which has been widely used as a benchmark for process control. The complete optimization structure for the styrene process including disturbances rejection is developed. The simulation results show the robustness of the proposed strategy and the capability to deal with disturbances while the economic objective is optimized.

  8. CADMIUM-RARE EARTH BORATE GLASS AS REACTOR CONTROL MATERIAL

    Science.gov (United States)

    Ploetz, G.L.; Ray, W.E.

    1958-11-01

    A reactor control rod fabricated from a cadmiumrare earth-borate glass is presented. The rare earth component of this glass is selected from among those rare earths having large neutron capture cross sections, such as samarium, gadolinium or europium. Partlcles of this glass are then dispersed in a metal matrix by standard powder metallurgy techniques.

  9. Keeping control when cutting through a reactor vessel

    International Nuclear Information System (INIS)

    Anon.

    1997-01-01

    UK Robotics' Advanced Teleoperation Controller (ATC) is a key component of one of the most extensive remote handling operations currently being undertaken - the removal of 165 mm diameter, 90 mm thick samples of carbon-manganese steel from the base of the Trawsfyndd reactor pressure vessel. These will then be used to assess the material properties of the vessel welds. (author)

  10. Application of fuzzy logic in the control of polymerization reactors

    NARCIS (Netherlands)

    Roffel, B.; Chin, P.A.

    1993-01-01

    Polymer reactors are ideal candidates for the application of fuzzy control. Many polymerization reactions are difficult to model, process measurements are often only available from laboratory analysis at infrequent time intervals and trace impurities can have a marked effect on the reaction. All

  11. Artificial Intelligent Control for a Novel Advanced Microwave Biodiesel Reactor

    International Nuclear Information System (INIS)

    Wali, W A; Hassan, K H; Cullen, J D; Al-Shamma'a, A I; Shaw, A; Wylie, S R

    2011-01-01

    Biodiesel, an alternative diesel fuel made from a renewable source, is produced by the transesterification of vegetable oil or fat with methanol or ethanol. In order to control and monitor the progress of this chemical reaction with complex and highly nonlinear dynamics, the controller must be able to overcome the challenges due to the difficulty in obtaining a mathematical model, as there are many uncertain factors and disturbances during the actual operation of biodiesel reactors. Classical controllers show significant difficulties when trying to control the system automatically. In this paper we propose a comparison of artificial intelligent controllers, Fuzzy logic and Adaptive Neuro-Fuzzy Inference System(ANFIS) for real time control of a novel advanced biodiesel microwave reactor for biodiesel production from waste cooking oil. Fuzzy logic can incorporate expert human judgment to define the system variables and their relationships which cannot be defined by mathematical relationships. The Neuro-fuzzy system consists of components of a fuzzy system except that computations at each stage are performed by a layer of hidden neurons and the neural network's learning capability is provided to enhance the system knowledge. The controllers are used to automatically and continuously adjust the applied power supplied to the microwave reactor under different perturbations. A Labview based software tool will be presented that is used for measurement and control of the full system, with real time monitoring.

  12. Artificial Intelligent Control for a Novel Advanced Microwave Biodiesel Reactor

    Science.gov (United States)

    Wali, W. A.; Hassan, K. H.; Cullen, J. D.; Al-Shamma'a, A. I.; Shaw, A.; Wylie, S. R.

    2011-08-01

    Biodiesel, an alternative diesel fuel made from a renewable source, is produced by the transesterification of vegetable oil or fat with methanol or ethanol. In order to control and monitor the progress of this chemical reaction with complex and highly nonlinear dynamics, the controller must be able to overcome the challenges due to the difficulty in obtaining a mathematical model, as there are many uncertain factors and disturbances during the actual operation of biodiesel reactors. Classical controllers show significant difficulties when trying to control the system automatically. In this paper we propose a comparison of artificial intelligent controllers, Fuzzy logic and Adaptive Neuro-Fuzzy Inference System(ANFIS) for real time control of a novel advanced biodiesel microwave reactor for biodiesel production from waste cooking oil. Fuzzy logic can incorporate expert human judgment to define the system variables and their relationships which cannot be defined by mathematical relationships. The Neuro-fuzzy system consists of components of a fuzzy system except that computations at each stage are performed by a layer of hidden neurons and the neural network's learning capability is provided to enhance the system knowledge. The controllers are used to automatically and continuously adjust the applied power supplied to the microwave reactor under different perturbations. A Labview based software tool will be presented that is used for measurement and control of the full system, with real time monitoring.

  13. Method of controlling power distribution in FBR type reactors

    International Nuclear Information System (INIS)

    Sawada, Shusaku; Kaneto, Kunikazu.

    1982-01-01

    Purpose: To attain the power distribution flattening with ease by obtaining a radial power distribution substantially in a constant configuration not depending on the burn-up cycle. Method: As the fuel burning proceeds, the radial power distribution is effected by the accumulation of fission products in the inner blancket fuel assemblies which varies the effect thereof as the neutron absorbing substances. Taking notice of the above fact, the power distribution is controlled in a heterogeneous FBR type reactor by varying the core residence period of the inner blancket assemblies in accordance with the charging density of the inner blancket assemblies in the reactor core. (Kawakami, Y.)

  14. Nonlinear dynamics and control of a recycle fixed bed reactor

    DEFF Research Database (Denmark)

    Recke, Bodil; Jørgensen, Sten Bay

    1997-01-01

    The purpose of this paper is twofold. Primarily to describe the dynamic behaviour that can be observed in a fixed bed reactor with recycle of unconverted reactant. Secondly to describe the possibilities of model reduction in order to facilitate control design. Reactant recycle has been shown...... to introduce periodic solution to the fixed bed reactor, a phenomenon which is not seen for the system without the recycle, at least not within the Peclet number range investigated in the present work. The possibility of model reduction by the methods of modal decomposition, and by characteristics...

  15. Nuclear reactor internals with control elements guides

    International Nuclear Information System (INIS)

    Baujat, J.; Chevereau, G.

    1991-01-01

    The internals have a lower plate, a superior plate, support columns and guide tubes for the control rods displacements. The lower section of the control rod guide tube have a base that fits into a bevelled seat in the lower plate. The guide tube is held into the seat by a spring, compressed between the base of the upper section of the tube and the lower plate

  16. Enabling autonomous control for space reactor power systems

    International Nuclear Information System (INIS)

    Wood, R. T.

    2006-01-01

    The application of nuclear reactors for space power and/or propulsion presents some unique challenges regarding the operations and control of the power system. Terrestrial nuclear reactors employ varying degrees of human control and decision-making for operations and benefit from periodic human interaction for maintenance. In contrast, the control system of a space reactor power system (SRPS) employed for deep space missions must be able to accommodate unattended operations due to communications delays and periods of planetary occlusion while adapting to evolving or degraded conditions with no opportunity for repair or refurbishment. Thus, a SRPS control system must provide for operational autonomy. Oak Ridge National Laboratory (ORNL) has conducted an investigation of the state of the technology for autonomous control to determine the experience base in the nuclear power application domain, both for space and terrestrial use. It was found that control systems with varying levels of autonomy have been employed in robotic, transportation, spacecraft, and manufacturing applications. However, autonomous control has not been implemented for an operating terrestrial nuclear power plant nor has there been any experience beyond automating simple control loops for space reactors. Current automated control technologies for nuclear power plants are reasonably mature, and basic control for a SRPS is clearly feasible under optimum circumstances. However, autonomous control is primarily intended to account for the non optimum circumstances when degradation, failure, and other off-normal events challenge the performance of the reactor and near-term human intervention is not possible. Thus, the development and demonstration of autonomous control capabilities for the specific domain of space nuclear power operations is needed. This paper will discuss the findings of the ORNL study and provide a description of the concept of autonomy, its key characteristics, and a prospective

  17. Graphics and control for in-reactor operations

    International Nuclear Information System (INIS)

    Smith, A.L.

    1996-01-01

    A wide range of manipulator systems has been developed to carry out remotely operated inspection, repair and maintenance tasks at the Magnox reactors in the United Kingdom. A key factor in the improvement of these systems in recent years has been the extensive use of computer graphics as a real-time aid to the manipulator operator. This is exemplified by the reactor pressure vessel inspection work at the Bradwell reactor which is described in detail. The graphics sub-system of the control system for the manipulator plays a unique and wide-ranging role. The 3D modelling and simulation capability of the IGRIP software has contributed to the conceptual design, detailed path planning, rehearsal support, public relations, real-time manipulator display, post inspection documentation and quality assurance. (UK)

  18. Adaptive control method for core power control in TRIGA Mark II reactor

    Science.gov (United States)

    Sabri Minhat, Mohd; Selamat, Hazlina; Subha, Nurul Adilla Mohd

    2018-01-01

    The 1MWth Reactor TRIGA PUSPATI (RTP) Mark II type has undergone more than 35 years of operation. The existing core power control uses feedback control algorithm (FCA). It is challenging to keep the core power stable at the desired value within acceptable error bands to meet the safety demand of RTP due to the sensitivity of nuclear research reactor operation. Currently, the system is not satisfied with power tracking performance and can be improved. Therefore, a new design core power control is very important to improve the current performance in tracking and regulate reactor power by control the movement of control rods. In this paper, the adaptive controller and focus on Model Reference Adaptive Control (MRAC) and Self-Tuning Control (STC) were applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, adaptive controller model, and control rods selection programming. The mathematical models of the reactor core were based on point kinetics model, thermal hydraulic models, and reactivity models. The adaptive control model was presented using Lyapunov method to ensure stable close loop system and STC Generalised Minimum Variance (GMV) Controller was not necessary to know the exact plant transfer function in designing the core power control. The performance between proposed adaptive control and FCA will be compared via computer simulation and analysed the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

  19. Research on digital PID control algorithm for HPCT

    International Nuclear Information System (INIS)

    Zeng Yi; Li Rui; Shen Tianjian; Ke Xinhua

    2009-01-01

    Digital PID applied in high-precision HPCT (High-precision current transducer) based on Digital Signal Processor (DSP) TMS320F2812 and special D/A converter was researched. By using increment style PID Control algorithm, the stability and precision of high-precision HPCT output voltage is improved. On basis of deeply analysing incremental digital PID, the scheme model of HPCT is proposed, the feasibility simulation using Matlab is given. Practical hardware circuit verified the incremental PID has closed-loop control process in tracking HPCT output voltage. (authors)

  20. Fabrication of control rods for the High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Sease, J.D.

    1998-01-01

    The High Flux Isotope Reactor (HFIR) is a research-type nuclear reactor that was designed and built in the early 1960s and has been in continuous operation since its initial criticality in 1965. Under current plans, the HFIR is expected to continue in operation until 2035. This report updates ORNL/TM-9365, Fabrication Procedure for HFIR Control Plates, which was mainly prepared in the early 1970's but was not issued until 1984, and reflects process changes, lessons learned in the latest control rod fabrication campaign, and suggested process improvements to be considered in future campaigns. Most of the personnel involved with the initial development of the processes and in part campaigns have retired or will retire soon. Because their unlikely availability in future campaigns, emphasis has been placed on providing some explanation of why the processes were selected and some discussions about the importance of controlling critical process parameters. Contained in this report is a description of the function of control rods in the reactor, the brief history of the development of control rod fabrication processes, and a description of procedures used in the fabrication of control rods. A listing of the controlled documents and procedures used in the last fabrication campaigns is referenced in Appendix A

  1. Fabrication of control rods for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sease, J.D.

    1998-03-01

    The High Flux Isotope Reactor (HFIR) is a research-type nuclear reactor that was designed and built in the early 1960s and has been in continuous operation since its initial criticality in 1965. Under current plans, the HFIR is expected to continue in operation until 2035. This report updates ORNL/TM-9365, Fabrication Procedure for HFIR Control Plates, which was mainly prepared in the early 1970's but was not issued until 1984, and reflects process changes, lessons learned in the latest control rod fabrication campaign, and suggested process improvements to be considered in future campaigns. Most of the personnel involved with the initial development of the processes and in part campaigns have retired or will retire soon. Because their unlikely availability in future campaigns, emphasis has been placed on providing some explanation of why the processes were selected and some discussions about the importance of controlling critical process parameters. Contained in this report is a description of the function of control rods in the reactor, the brief history of the development of control rod fabrication processes, and a description of procedures used in the fabrication of control rods. A listing of the controlled documents and procedures used in the last fabrication campaigns is referenced in Appendix A.

  2. Boiling water reactor radiation shielded Control Rod Drive Housing Supports

    Energy Technology Data Exchange (ETDEWEB)

    Baversten, B.; Linden, M.J. [ABB Combustion Engineering Nuclear Operations, Windsor, CT (United States)

    1995-03-01

    The Control Rod Drive (CRD) mechanisms are located in the area below the reactor vessel in a Boiling Water Reactor (BWR). Specifically, these CRDs are located between the bottom of the reactor vessel and above an interlocking structure of steel bars and rods, herein identified as CRD Housing Supports. The CRD Housing Supports are designed to limit the travel of a Control Rod and Control Rod Drive in the event that the CRD vessel attachement went to fail, allowing the CRD to be ejected from the vessel. By limiting the travel of the ejected CRD, the supports prevent a nuclear overpower excursion that could occur as a result of the ejected CRD. The Housing Support structure must be disassembled in order to remove CRDs for replacement or maintenance. The disassembly task can require a significant amount of outage time and personnel radiation exposure dependent on the number and location of the CRDs to be changed out. This paper presents a way to minimize personal radiation exposure through the re-design of the Housing Support structure. The following paragraphs also delineate a method of avoiding the awkward, manual, handling of the structure under the reactor vessel during a CRD change out.

  3. Control rod homogenization in heterogeneous sodium-cooled fast reactors

    International Nuclear Information System (INIS)

    Andersson, Mikael

    2016-01-01

    The sodium-cooled fast reactor is one of the candidates for a sustainable nuclear reactor system. In particular, the French ASTRID project employs an axially heterogeneous design, proposed in the so-called CFV (low sodium effect) core, to enhance the inherent safety features of the reactor. This thesis focuses on the accurate modeling of the control rods, through the homogenization method. The control rods in a sodium-cooled fast reactor are used for reactivity compensation during the cycle, power shaping, and to shutdown the reactor. In previous control rod homogenization procedures, only a radial description of the geometry was implemented, hence the axially heterogeneous features of the CFV core could not be taken into account. This thesis investigates the different axial variations the control rod experiences in a CFV core, to determine the impact that these axial environments have on the control rod modeling. The methodology used in this work is based on previous homogenization procedures, the so-called equivalence procedure. The procedure was newly implemented in the PARIS code system in order to be able to use 3D geometries, and thereby be take axial effects into account. The thesis is divided into three parts. The first part investigates the impact of different neutron spectra on the homogeneous control-rod cross sections. The second part investigates the cases where the traditional radial control-rod homogenization procedure is no longer applicable in the CFV core, which was found to be 5-10 cm away from any material interface. In the third part, based on the results from the second part, a 3D model of the control rod is used to calculate homogenized control-rod cross sections. In a full core model, a study is made to investigate the impact these axial effects have on control rod-related core parameters, such as the control rod worth, the capture rates in the control rod, and the power in the adjacent fuel assemblies. All results were compared to a Monte

  4. The design and construction of a controllable reactor with a HTS control winding

    International Nuclear Information System (INIS)

    Wass, Torbjoern; Hoernfeldt, Sven; Valdemarsson, Stefan

    2006-01-01

    Reactive power compensation is vital for obtaining efficient operation of long transmission power lines or cables. The need of reactive power changes with the load of the transmission line. Discrete units of conventional reactors are therefore switched in and out in order to obtain more efficient reactive power compensation. A continuous reactive compensation will reduce the transmission losses and increase the transmission capacity of active power. We have designed and constructed a one phase small scale prototype of a controllable shunt reactor with a high temperature superconducting control winding. The reactor consists basically of two windings and an iron core. The control winding is placed so that it generates a DC magnetic field perpendicular to the main AC magnetic field. Thus the DC current in the control winding can control the direction of the magnetization of the iron core and thereby the reactance of the reactor. Such a control winding will have low losses and give the reactor a large dynamic range. For this small scale reactor we found that the reactive power could be varied with a factor six. We have demonstrated the feasibility to design large scale controllable shunt reactors with large dynamic range and low losses utilizing a control winding made of a high temperature superconductor

  5. Reactivity control system of the high temperature engineering test reactor

    International Nuclear Information System (INIS)

    Tachibana, Yukio; Sawahata, Hiroaki; Iyoku, Tatsuo; Nakazawa, Toshio

    2004-01-01

    The reactivity control system of the high temperature engineering test reactor (HTTR) consists of a control rod system and a reserve shutdown system. During normal operation, reactivity is controlled by the control rod system, which consists of 32 control rods (16 pairs) and 16 control rod drive mechanisms except for the case when the center control rods are removed to perform an irradiation test. In an unlikely event that the control rods fail to be inserted, reserve shutdown system is provided to insert pellets of neutron-absorbing material into the core. Alloy 800H is chosen for the metallic parts of the control rods. Because the maximum temperature of the control rods reaches about 900 deg. C at reactor scrams, structural design guideline and design material data on Alloy 800H are needed for the high temperature design. The design guideline for the HTTR control rod is based on ASME Code Case N-47-21. Design material data is also determined and shown in this paper. Observing the guideline, temperature and stress analysis were conducted; it can be confirmed that the target life of the control rods of 5 years can be achieved. Various tests conducted for the control rod system and the reserve shutdown system are also described

  6. Control rod drive mechanism with shock absorber for nuclear reactor

    International Nuclear Information System (INIS)

    Chevereau, G.

    1989-01-01

    The mechanism usable in a PWR has a shaft carrying the bar vertically displaceable in the reactor internals and a dash pot with a hydraulic cylinder and a piston. The cylinder has a large diameter perforated upper section to the cylinder, a small diameter lower section, a piston traversed by the control rod sized to fit into the upper section and forced downwards when the control descends. The shock absorbing chamber is defined between the piston and the upper section [fr

  7. Autonomous Control Capabilities for Space Reactor Power Systems

    International Nuclear Information System (INIS)

    Wood, Richard T.; Neal, John S.; Brittain, C. Ray; Mullens, James A.

    2004-01-01

    The National Aeronautics and Space Administration's (NASA's) Project Prometheus, the Nuclear Systems Program, is investigating a possible Jupiter Icy Moons Orbiter (JIMO) mission, which would conduct in-depth studies of three of the moons of Jupiter by using a space reactor power system (SRPS) to provide energy for propulsion and spacecraft power for more than a decade. Terrestrial nuclear power plants rely upon varying degrees of direct human control and interaction for operations and maintenance over a forty to sixty year lifetime. In contrast, an SRPS is intended to provide continuous, remote, unattended operation for up to fifteen years with no maintenance. Uncertainties, rare events, degradation, and communications delays with Earth are challenges that SRPS control must accommodate. Autonomous control is needed to address these challenges and optimize the reactor control design. In this paper, we describe an autonomous control concept for generic SRPS designs. The formulation of an autonomous control concept, which includes identification of high-level functional requirements and generation of a research and development plan for enabling technologies, is among the technical activities that are being conducted under the U.S. Department of Energy's Space Reactor Technology Program in support of the NASA's Project Prometheus. The findings from this program are intended to contribute to the successful realization of the JIMO mission

  8. Comparative analysis of nuclear reactor control system designs

    International Nuclear Information System (INIS)

    Russcher, G.E.

    1975-01-01

    Control systems are vital to the safe operation of nuclear reactors. Their seismic design requirements are some of the most important criteria governing reactor system design evaluation. Consequently, the seismic analysis for nuclear reactors is directed to include not only the mechanical and structural seismic capabilities of a reactor, but the control system functional requirements as well. In the study described an alternate conceptual design of a safety rod system was compared with a prototypic system design to assess their relative functional reliabilities under design seismic conditions. The comparative methods utilized standard success tree and decision tree techniques to determine the relative figures of merit. The study showed: (1) The methodology utilized can provide both qualitative and quantitative bases for design decisions regarding seismic functional capabilities of two systems under comparison, (2) the process emphasizes the visibility of particular design features that are subject to common mode failure while under seismic loading, and (3) minimal improvement was shown to be available in overall system seismic performance of an independent conceptual design, however, it also showed the system would be subject to a new set of operational uncertainties which would have to be resolved by extensive development programs

  9. Method of controlling the water quality in nuclear reactors

    International Nuclear Information System (INIS)

    Ibe, Hidefumi.

    1985-01-01

    Purpose: To obtain a simple and reliable water quality calculation system and water quality control method based thereon for the entire primary coolant circuits in BWR type reactors. Method: In a method of controlling the water quality of the reactor water by injecting hydrogen into the primary coolant circuits of a nuclear reactor, by utilizing a first linear relationship established between the concentration of oxygen and hydrogen in the main steam system and the concentration of radiolysis products in the reactor core and separators and mixing plenum portions, each of the above-mentioned concentrations is calculated from the concentrations for hydrogen or oxygen. Further, by utilizing the first linear relationship established between the concentrations for the oxygen and hydrogen in the recycling system and the concentration of the radiolysis products in the system from the downcomer to the lower plenum portion, the above-mentioned concentration is calculated from the concentration for oxygen and hydrogen. Then, the hydrogen injection rate into the primary coolant system is determined such that the calculated value takes an aimed value. (Ikeda, J.)

  10. Control device for start-up of reactor depressurization system

    International Nuclear Information System (INIS)

    Suzuki, Hiroshi; Saito, Minoru; Oda, Shingo; Miura, Satoshi; Hashimoto, Koji; Tate, Hitoshi; Fujii, Kazunobu

    1998-01-01

    The present invention concerns are emergency reactor core cooling system (ECCS) of a BWR type reactor and provides a control device for start-up of an automatic depressurization system. Namely, the device has an object of preventing erroneous opening of a main steam escape safety value when testing a start-up signal circuit of an automatic depressurization system for testing the automatic depressurization system. A start-up signal circuit receives both signals of a reactor container pressure high signal and a reactor pressure vessel water level low signal and outputs an automatic start-up signal for compulsorily opening a main steam escape safety valve automatically. A test switch having a self-holding circuit is disposed to a central control chamber. A test signal circuit is disposed for preventing transfer of an erroneous start-up signal to the main steam escape safety valve due to a simulation signal during output test signals by the test switch. (I.S.)

  11. Human machine interface for research reactor instrumentation and control system

    International Nuclear Information System (INIS)

    Mohd Sabri Minhat; Mohd Idris Taib; Izhar Abu Hussin; Zareen Khan Abdul Jalil Khan; Nurfarhana Ayuni Joha

    2010-01-01

    Most present design of Human Machine Interface for Research Reactor Instrumentation and Control System is modular-based, comprise of several cabinets such as Reactor Protection System, Control Console, Information Console as well as Communication Console. The safety, engineering and human factor will be concerned for the design. Redundancy and separation of signal and power supply are the main factor for safety consideration. The design of Operator Interface absolutely takes consideration of human and environmental factors. Physical parameters, experiences, trainability and long-established habit patterns are very important for user interface, instead of the Aesthetic and Operator-Interface Geometry. Physical design for New Instrumentation and Control System of RTP are proposed base on the state-of- the-art Human Machine Interface design. (author)

  12. Replacement of the Advanced Test Reactor control room

    International Nuclear Information System (INIS)

    Durney, J.L.; Klingler, W.B.

    1990-01-01

    The control room for the Advanced Test Reactor has been replaced to provide modern equipment utilizing current standards and meeting the current human factors requirements. The control room was designed in the early 1960 era and had not been significantly upgraded since the initial installation. The replacement did not change any of the safety circuits or equipment but did result in replacement of some of the recorders that display information from the safety systems. The replacement was completed in concert with the replacement of the control room simulator which provided important feedback on the design. The design successfully incorporates computer-based systems into the display of the plant variables. This improved design provides the operator with more information in a more usable form than was provided by the original design. The replacement was successfully completed within the scheduled time thereby minimizing the down time for the reactor

  13. Drive mechanism nuclear reactor control rod

    International Nuclear Information System (INIS)

    Brooks, J.G. Jr.; Maure, D.R.; Meijer, C.H.

    1978-01-01

    An improved method and apparatus for operating magnetic stepping-type mechanisms. The current flowing in the coils of magnetic stepping-type mechanisms of the kind, for instance, that are used in control-element drive mechanisms is sensed and used to monitor operation of the mechanism. Current waveforms that characterize the motion of the mechanism are used to trigger changes in drive voltage and to verify that the drive mechanism is operating properly. In addition, incipient failures are detected through the observation of differences between the observed waveform and waveforms that characterize proper operation

  14. Burn Control in Fusion Reactors via Nonlinear Stabilization Techniques

    International Nuclear Information System (INIS)

    Schuster, Eugenio; Krstic, Miroslav; Tynan, George

    2003-01-01

    Control of plasma density and temperature magnitudes, as well as their profiles, are among the most fundamental problems in fusion reactors. Existing efforts on model-based control use control techniques for linear models. In this work, a zero-dimensional nonlinear model involving approximate conservation equations for the energy and the densities of the species was used to synthesize a nonlinear feedback controller for stabilizing the burn condition of a fusion reactor. The subignition case, where the modulation of auxiliary power and fueling rate are considered as control forces, and the ignition case, where the controlled injection of impurities is considered as an additional actuator, are treated separately.The model addresses the issue of the lag due to the finite time for the fresh fuel to diffuse into the plasma center. In this way we make our control system independent of the fueling system and the reactor can be fed either by pellet injection or by puffing. This imposed lag is treated using nonlinear backstepping.The nonlinear controller proposed guarantees a much larger region of attraction than the previous linear controllers. In addition, it is capable of rejecting perturbations in initial conditions leading to both thermal excursion and quenching, and its effectiveness does not depend on whether the operating point is an ignition or a subignition point.The controller designed ensures setpoint regulation for the energy and plasma parameter β with robustness against uncertainties in the confinement times for different species. Hence, the controller can increase or decrease β, modify the power, the temperature or the density, and go from a subignition to an ignition point and vice versa

  15. Method of nuclear reactor control using a variable temperature load dependent set point

    International Nuclear Information System (INIS)

    Kelly, J.J.; Rambo, G.E.

    1982-01-01

    A method and apparatus for controlling a nuclear reactor in response to a variable average reactor coolant temperature set point is disclosed. The set point is dependent upon percent of full power load demand. A manually-actuated ''droop mode'' of control is provided whereby the reactor coolant temperature is allowed to drop below the set point temperature a predetermined amount wherein the control is switched from reactor control rods exclusively to feedwater flow

  16. Robust nonlinear control of nuclear reactors under model uncertainty

    International Nuclear Information System (INIS)

    Park, Moon Ghu

    1993-02-01

    A nonlinear model-based control method is developed for the robust control of a nuclear reactor. The nonlinear plant model is used to design a unique control law which covers a wide operating range. The robustness is a crucial factor for the fully automatic control of reactor power due to time-varying, uncertain parameters, and state estimation error, or unmodeled dynamics. A variable structure control (VSC) method is introduced which consists of an adaptive performance specification (fime control) after the tracking error reaches the narrow boundary-layer by a time-optimal control (coarse control). Variable structure control is a powerful method for nonlinear system controller design which has inherent robustness to parameter variations or external disturbances using the known uncertainty bounds, and it requires very low computational efforts. In spite of its desirable properties, conventional VSC presents several important drawbacks that limit its practical applicability. One of the most undesirable phenomena is chattering, which implies extremely high control activity and may excite high-frequency unmodeled dynamics. This problem is due to the neglected actuator time-delay or sampling effects. The problem was partially remedied by replacing chattering control by a smooth control inter-polation in a boundary layer neighnboring a time-varying sliding surface. But, for the nuclear reactor systems which has very fast dynamic response, the sampling effect may destroy the narrow boundary layer when a large uncertainty bound is used. Due to the very short neutron life time, large uncertainty bound leads to the high gain in feedback control. To resolve this problem, a derivative feedback is introduced that gives excellent performance by reducing the uncertainty bound. The stability of tracking error dynamics is guaranteed by the second method of Lyapunov using the two-level uncertainty bounds that are obtained from the knowledge of uncertainty bound and the estimated

  17. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  18. Reactor physics calculations for the control of the advanced neutron source reactor

    International Nuclear Information System (INIS)

    Difilippo, F.C.; Abu-Shehadeh, M.; Perez, R.B.

    1988-01-01

    Efficient production of extremely high fluxes requires compact cores with consequent high power densities and initial excess reactivities. Strong space dependent neutron spectras and limited access to the small core are other characteristics that make design of the control system of these type of facilities an interesting problem. We present calculations of the worths of 10 B to reduce the initial excess reactivity, the worth of Hf and B control rods, and the neutron lifetimes, for the case of candidate designs for the Advanced Neutron Source reactor. 4 refs., 4 figs., 2 tabs

  19. Digital IP Protection Using Threshold Voltage Control

    OpenAIRE

    Davis, Joseph; Kulkarni, Niranjan; Yang, Jinghua; Dengi, Aykut; Vrudhula, Sarma

    2016-01-01

    This paper proposes a method to completely hide the functionality of a digital standard cell. This is accomplished by a differential threshold logic gate (TLG). A TLG with $n$ inputs implements a subset of Boolean functions of $n$ variables that are linear threshold functions. The output of such a gate is one if and only if an integer weighted linear arithmetic sum of the inputs equals or exceeds a given integer threshold. We present a novel architecture of a TLG that not only allows a single...

  20. Development of an Advanced Digital Reactor Protection System Using Diverse Dual Processors to Prevent Common-Mode Failure

    International Nuclear Information System (INIS)

    Shin, Hyun Kook; Nam, Sang Ku; Sohn, Se Do; Chang, Hoon Seon

    2003-01-01

    The advanced digital reactor protection system (ADRPS) with diverse dual processors has been developed to prevent common-mode failure (CMF). The principle of diversity is applied to both hardware design and software design. For hardware diversity, two different types of CPUs are used for the bistable processor and local coincidence logic (LCL) processor. The Versa Module Eurocard-based single board computers are used for the CPU hardware platforms. The QNX operating system and the VxWorks operating system were selected for software diversity. Functional diversity is also applied to the input and output modules, and to the algorithm in the bistable processors and LCL processors. The characteristics of the newly developed digital protection system are described together with the preventive capability against CMF. Also, system reliability analysis is discussed. The evaluation results show that the ADRPS has a good preventive capability against the CMF and is a highly reliable reactor protection system

  1. Development and study of a control and reactor shutdown device for FBR-type reactors with a modified open core

    International Nuclear Information System (INIS)

    Goswami, S.

    1983-01-01

    The doctoral thesis at hand presents a newly designed control and shutdown device to be used for output control and fast shutdown of modified open core FBR-type reactors. The task was the design of a new control and shutdown device having economic and operation advantages, using reactor components time-tested under reactor conditions. This control and shutdown device was adapted to the specific needs concerning dimensions and design. The actuation is based on the magnetic-jack principle, which has been upgraded for the purpose. The principle is now combined with pneumatic acceleration. The improvements mainly concern a smaller number of piece parts and system simplification. (orig./RW) [de

  2. Application of H∞ control theory to power control of a nonlinear reactor model

    International Nuclear Information System (INIS)

    Suzuki, Katsuo; Shimazaki, Junya; Shinohara, Yoshikuni

    1993-01-01

    The H∞ control theory is applied to the compensator design of a nonlinear nuclear reactor model, and the results are compared with standard linear quadratic Gaussian (LQG) control. The reactor model is assumed to be provided with a control rod drive system having the compensation of rod position feedback. The nonlinearity of the reactor model exerts a great influence on the stability of the control system, and hence, it is desirable for a power control system of a nuclear reactor to achieve robust stability and to improve the sensitivity of the feedback control system. A computer simulation based on a power control system synthesized by LQG control was performed revealing that the control system has some stationary offset and less stability. Therefore, here, attention is given to the development of a methodology for robust control that can withstand exogenous disturbances and nonlinearity in view of system parameter changes. The developed methodology adopts H∞ control theory in the feedback system and shows interesting features of robustness. The results of the computer simulation indicate that the feedback control system constructed by the developed H∞ compensator possesses sufficient robustness of control on the stability and disturbance attenuation, which are essential for the safe operation of a nuclear reactor

  3. The use of fuzzy logic control in solving problems related to the control of nuclear research reactors

    International Nuclear Information System (INIS)

    EL-Badawy, O.I.

    1998-01-01

    In this thesis, the fuzzy logic controlling technique was used in nuclear engineering technology, especially application to control the nuclear research reactors. It also shows the main idea of fuzzy logic control techniques and uses one of them for the implementation of the first Egyptian research reactor reactor ETRR-1 as the case of research reactors to control not only the reactor power, but also upgrading its control function; such as automation of the power compensation via moving the manual control rods, and controlling the core water temperature limits taking into account the safety design rules to avoid rules to avoid shutdown or scram the reactor. Finally a package of software program was developed to simulate and implement the fuzzy logic controller technique for the following applications: 1- Reactor power control. 2- Automatic power compensation control. 3- The core water temperature control

  4. Electromotor control rod drive for nuclear reactors

    International Nuclear Information System (INIS)

    Baker, S.M.

    1975-01-01

    The positioning of a control rod arranged in a pressure vessel takes place with a drive. This protrudes out of the pressure vessel through a support and is formed from a rotating field motor with energy source, e.g. alternating current connection. Its stator surrounds a section of a pressure casing which covers the length of the drive. The rotor is arranged in the pressure casing and interacts with a shaft lying in the rotation axis. Furthermore, segments are hinged on it, each of which forms two arms of a rocker. Each segment can be revolved against a storing force in a plane containing the rotation axis, through the stator field acting on one of the rocker arms. In order that the drive motor is automatically blocked should the electricity supply fail, the other rocker arm can be connected with a fixed cased component of the drive having the effect of a friction break or a form-locking mechanical catch. (DG/LH) [de

  5. Monitoring and control of the Rossendorf research reactor using a microcomputerized automation system

    International Nuclear Information System (INIS)

    Ba weg, F.; Enkelmann, W.; Klebau, J.

    1982-01-01

    A decentral hierarchic information system (HIS) is presented, which has been developed for monitoring and control of the Rossendorf Research Reactor RFR, but which may also be considered the prototype of a digital automation system (AS) to be used in power stations. The functions integrated in the HIS are as follows: process monitoring, process control, and use of a specialized industrial robot for control of charging and discharging of the materials to be irradiated. The AS is realized on the basis of the process computer system PRA 30 (A 6492) developed in the GDR and including a computer K 1630 and the intelligent process terminals ursadat 5000 connected by a fast serial interface (IFLS). (author)

  6. Test results of the reactor inlet coolant temperature control system of HTTR

    International Nuclear Information System (INIS)

    Saito, Kenji; Nakagawa, Shigeaki; Hirato, Yoji

    2004-04-01

    The reactor control system of HTTR is composed of the reactor power control system, the reactor inlet coolant temperature control system, the primary coolant flow rate control system and so on. The reactor control system of HTTR achieves reactor power 30 MW, reactor outlet coolant temperature 850degC, reactor inlet coolant temperature 395degC under the condition that primary coolant flow rate is fixed. In the Rise-to-Power Test, the performance test of the reactor inlet coolant temperature control system was carried out in order to confirm the control capability of this control system. This report shows the test results of performance test. As a result, the control parameters, which can control the reactor inlet coolant temperature stably during the reactor operation, were successfully selected. And it was confirmed that the reactor inlet coolant temperature control system has the capability of controlling the reactor inlet coolant temperature stably against any disturbances on the basis of operational condition of HTTR. (author)

  7. Dimensional control and check of field machining parts for reactor internals installation

    International Nuclear Information System (INIS)

    Zhang Caifang

    2010-01-01

    Some key issues of dimensional control for reactor internals installation are analyzed, and important technical requirements of crucial quality control elements on the measurement, machining, and checking of reactor internals filed machining parts are discussed. Moreover, provisions on quality control and risk prevention of reactor internals filed machining parts are presented in this paper. (author)

  8. Multivariable robust control of an integrated nuclear power reactor

    Directory of Open Access Journals (Sweden)

    A. Etchepareborda

    2002-12-01

    Full Text Available The design of the main control system of the CAREM nuclear power plant is presented. This plant is an inherently safe low-power nuclear reactor with natural convection on the primary coolant circuit and is self-pressurized with a steam dome on the top of the pressure vessel (PV. It is an integrated reactor as the whole primary coolant circuit is within the PV. The primary circuit transports the heat to the secondary circuit through once-through steam generators (SG. There is a feedwater valve at the inlet of the SG and a turbine valve at the outlet of the SG. The manipulated variables are the aperture of these valves and the reactivity of the control rods. The control target is to regulate the primary and secondary pressures and to monitor steam flow reference ramps on a range of nominal flow from 100% to 40%. The requirements for the control system are robust stability, low-order simple controllers and transient/permanent error bounding. The controller design is based on a detailed RETRAN plant model, from which linear perturbed open-loop dynamic models at different powers are identified. Two low-order nominal models with their associated uncertainties are chosen for two different power ranges. Robust controllers with acceptable performances are designed for each range. Numerical optimization based on the loop-shaping method is used for the controller design. The designed controllers are implemented in the RETRAN model and tested in simulations achieving successful results.

  9. Power control device for heavy water moderated reactor

    International Nuclear Information System (INIS)

    Matsushima, Hidesuke; Masuda, Hiroyuki.

    1978-01-01

    Purpose: To improve self controllability of a nuclear power plant, as well as enable continuous power level control by a controlled flow of moderators in void pipes provided in a reactor core. Constitution: Hollow void pipes are provided in a reactor core to which a heavy water recycle loop for power control, a heavy water recycle pump for power control, a heavy water temperature regulator and a heavy water flow rate control valve for power control are connected in series to constitute a heavy water recycle loop for flowing heavy water moderators. The void ratio in each of the void pipes are calculated by a process computer to determine the flow rate and the temperature for the recycled heavy water. Based on the above calculation result, the heavy water temperature regulator is actuated by way of a temperature setter at the heavy water inlet and the heavy water flow rate is controlled by the actuation of the heavy water flow rate control valve. (Kawakami, Y.)

  10. Qualification of safety-critical software for digital reactor safety system in nuclear power plants

    International Nuclear Information System (INIS)

    Kwon, Kee-Choon; Park, Gee-Yong; Kim, Jang-Yeol; Lee, Jang-Soo

    2013-01-01

    This paper describes the software qualification activities for the safety-critical software of the digital reactor safety system in nuclear power plants. The main activities of the software qualification processes are the preparation of software planning documentations, verification and validation (V and V) of the software requirements specifications (SRS), software design specifications (SDS) and codes, and the testing of the integrated software and integrated system. Moreover, the software safety analysis and software configuration management are involved in the software qualification processes. The V and V procedure for SRS and SDS contains a technical evaluation, licensing suitability evaluation, inspection and traceability analysis, formal verification, software safety analysis, and an evaluation of the software configuration management. The V and V processes for the code are a traceability analysis, source code inspection, test case and test procedure generation. Testing is the major V and V activity of the software integration and system integration phases. The software safety analysis employs a hazard operability method and software fault tree analysis. The software configuration management in each software life cycle is performed by the use of a nuclear software configuration management tool. Through these activities, we can achieve the functionality, performance, reliability, and safety that are the major V and V objectives of the safety-critical software in nuclear power plants. (author)

  11. Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

    International Nuclear Information System (INIS)

    Peterson, Per; Greenspan, Ehud

    2015-01-01

    This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m 3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m 3 . This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X-PREX facility uses novel

  12. Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering; Greenspan, Ehud [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering

    2015-02-09

    This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X

  13. Use of hafnium in control bars of nuclear reactors

    International Nuclear Information System (INIS)

    Ramirez S, J.R.; Alonso V, G.

    2003-01-01

    Recently the use of hafnium as neutron absorber material in nuclear reactors has been reason of investigation by virtue of that this material has nuclear properties as to the neutrons absorption and structural that can prolong the useful life of the control mechanisms of the nuclear reactors. In this work some of those more significant hafnium properties are presented like nuclear material. Also there are presented calculations carried out with the HELIOS code for fuel cells of uranium oxide and of uranium and plutonium mixed oxides under controlled conditions with conventional bars of boron carbide and also with similar bars to which are substituted the absorbent material by metallic hafnium, the results are presented in this work. (Author)

  14. Using a digital signal processor as a data stream controller for digital subtraction angiography

    International Nuclear Information System (INIS)

    Meng, J.D.; Katz, J.E.

    1991-10-01

    High speed, flexibility, and good arithmetic abilities make digital signal processors (DSP) a good choice as input/output controllers for real time applications. The DSP can be made to pre-process data in real time to reduce data volume, to open early windows on what is being acquired and to implement local servo loops. We present an example of a DSP as an input/output controller for a digital subtraction angiographic imaging system. The DSP pre-processes the raw data, reducing data volume by a factor of two, and is potentially capable of producing real-time subtracted images for immediate display

  15. Regulatory issues of digital instrumentation and control system in Lungmen project

    International Nuclear Information System (INIS)

    Chuang, C.F.; Chou, H.P.

    2004-01-01

    The Lungmen Nuclear Power Station (LNPS) is currently under construction in Taiwan, which consists of 2 advanced boiling water reactor (ABWR) units. The instrumentation and control (IC) systems of the LNPS are based on the state-of-the-art modernized fully integrated digital design. These IC systems possess many advantages and distinguished features comparing to traditional analog IC systems, they enjoy set-point stability, self-diagnostic and automatic testing ability, fault tolerance and avoidance, low power requirements, data handling and storage capability, as well as enhanced human-machine interfaces. This paper presents regulatory overviews, regulatory requirements, current major regulatory issues, as well as the areas of regulatory concerns and the lessons learned on the digital IC systems in the Lungmen Project

  16. Optimal filtering, parameter tracking, and control of nonlinear nuclear reactors

    International Nuclear Information System (INIS)

    March-Leuba, C.; March-Leuba, J.; Perez, R.B.

    1988-01-01

    This paper presents a new formulation of a class of nonlinear optimal control problems in which the system's signals are noisy and some system parameters are changing arbitrarily with time. The methodology is validated with an application to a nonlinear nuclear reactor model. A variational technique based on Pontryagin's Maximum Principle is used to filter the noisy signals, estimate the time-varying parameters, and calculate the optimal controls. The reformulation of the variational technique as an initial value problem allows this microprocessor-based algorithm to perform on-line filtering, parameter tracking, and control

  17. Quality indexes for selecting control materials of the nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Martinez-Val, J.M.; Pena, J.; Esteban Naudin, A.

    1981-01-01

    Quality indexes are established and valued for selecting control materials, The requirements for accomplishing such purposes are explained with detailed analysis: absortion cross section must be as high as possible, adequate reactivity evolution versus depletion, good resistance to radiation, appropiate thermal stability, mechanical resistance and ductility, chemical compatibility with the environment, good heat transfer properties, abundant in the nature and low costs. At present Westinghouse desire to commercialize hafnium as control material shows the exciting task of looking for new materials controlling nuclear reactors.

  18. Reactor

    International Nuclear Information System (INIS)

    Toyama, Masahiro; Kasai, Shigeo.

    1978-01-01

    Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

  19. System and method for air temperature control in an oxygen transport membrane based reactor

    Science.gov (United States)

    Kelly, Sean M

    2016-09-27

    A system and method for air temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

  20. System and method for temperature control in an oxygen transport membrane based reactor

    Science.gov (United States)

    Kelly, Sean M.

    2017-02-21

    A system and method for temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

  1. Reactor power control device in BWR power plant

    International Nuclear Information System (INIS)

    Kurosawa, Tsuneo.

    1997-01-01

    The present invention provides a device for controlling reactor power based on a start-up/shut down program in a BWR type reactor, as well as for detecting deviation, if occurs, of the power from the start-up/shut down program, to control a recycling flow rate control system or control rod drive mechanisms. Namely, a power instruction section successively executes the start-up/shut down program and controls the coolant recycling system and the control rod driving mechanisms to control the power. A current state monitoring and calculation section receives a process amount, calculates parameters showing the plant state, compares/monitors them with predetermined values, detecting the deviation, if occurs, of the plant state from the start-up/shut down program, and prevents output of a power increase control signal which leads to power increase. A forecasting and monitoring/calculation section forecasts and calculates the plant state when not yet executed steps of the start-up/shut down program are performed, stops the execution of the start-up/shut down program in the next step in a case of forecasting that the results of the calculation will deviate from the start-up/shut down program. (I.S.)

  2. SIMULACIÓN DE CONTROLADORES DIGITALES SIMULATION OF DIGITAL CONTROLLERS

    Directory of Open Access Journals (Sweden)

    Carlos Álvarez G

    2009-12-01

    Full Text Available El presente trabajo tiene como objetivo la implementación de controladores digitales en un entorno de simulación controlado, para esto se desarrolla una plataforma de hardware que permite ejecutar los programas en lenguaje C generados en una estación de trabajo. Estos programas corresponden al controlador y a la planta que son generados por un software que genera dichos programas a partir de sus parámetros de modelación aplicando teoría de control digital sobre procesos reales.This paper describes an implementation of digital controllers in a simulation environment for including a hardware platform for running programs generated on a workstation. These programs for both the controller and the plant are generated by software based on parameters using digital control theory for real processes.

  3. MAS2-8 radar and digital control unit

    Science.gov (United States)

    Oberg, J. M.; Ulaby, F. T.

    1974-01-01

    The design of the MAS 2-8 (2 to 8 GHz microwave-active spectrometer), a ground-based sensor system, is presented. A major modification in 1974 to the MAS 2-8, that of a control subsystem to automate the data-taking operation, is the prime focus. The digital control unit automatically changes all system parameters except FM rate and records the return signal on paper tape. The overall system operation and a detailed discussion of the design and operation of the digital control unit are presented.

  4. A digital method for period measurements in a nuclear reactor; Um metodo digital para medidas de periodo em um reator nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Mundim, Sergio Gorretta

    1971-02-15

    The present paper begins by giving a theoretical treatment for the nuclear reactor period. The conventional method of measuring the period is analysed and some previously developed digital methods are described. The paper criticises the latter, pointing out some deficiencies which the proposed process is able to eliminate. All errors connected with this process are also analysed. The paper presents suitable solutions to reduce them to a minimum. The total error is found to he less than the error presented by the other methods described. A digital period meter is designed with memory resources and an automatic scaler changer. Integrated circuits specifications are used in it. Real time experiments with nuclear reactors were made in order to check te validity of the method. The data acquired were applied to a simulated digital period meter implemented in a general purpose computer. The nuclear part of the work was developed at the 'Comissao Nacional de Energia Nuclear' and the simulation work was dane at the 'Departamento de Calculo Cientifico' of COPPE, which also advised the author in the completion of this thesis. (author)

  5. Design of Multi Objectives Control Systems to Control Nuclear Reactor Power

    International Nuclear Information System (INIS)

    Abdelaal, M.M.Z.

    2013-01-01

    The Egyptian Testing Research Reactor (ETRR-2) nonlinear twelfth order model is linearized and reduced to lower order model. Model order reduction methodologies such as balanced truncation, Schur reduction method, Hankel approximation and Coprime factorization have been used in the reduction process. The reactor actually controlled by PD controller with fixed tuning parameters. LMI state feedback, LMI-pool assignment, H ∞ and observer based controllers based third order model are proposed to be used in the reactor power control instead of the PD controller. A comparison of LMI, LMI-Pole placement,H ∞ control systems and those of based observer relative to the PD controller has been performed which showed better response and disturbance rejection for the proposed controllers.

  6. Reactor

    International Nuclear Information System (INIS)

    Ikeda, Masaomi; Kashimura, Kazuo; Inoue, Kazuyuki; Nishioka, Kazuya.

    1979-01-01

    Purpose: To facilitate the construction of a reactor containment building, whereby the inspections of the outer wall of a reactor container after the completion of the construction of the reactor building can be easily carried out. Constitution: In a reactor accommodated in a container encircled by a building wall, a space is provided between the container and the building wall encircling the container, and a metal wall is provided in the space so that it is fitted in the building wall in an attachable or detatchable manner. (Aizawa, K.)

  7. Applications of artificial intelligence to reactor and plant control

    International Nuclear Information System (INIS)

    Bernard, J.A.

    1989-01-01

    Potential improvements in plant efficiency and reliability are often cited as reasons for developing and applying artificial intelligence (AI) techniques, principally expert systems, to the control and operation of nuclear reactors. Nevertheless, there have been few such applications and then mostly at the prototype level. Therefore, if AI techniques are to contribute to process control, methods must be identified by which rule-based and analytic approaches can be merged. This hypothesis is the basic premise of this article. Presented below are 1. a brief review of the human approach towards process control, 2. a discussion of the suitability of AI methodologies for the performance of control tasks, 3. examples of AI applications to both open- and closed-loop control, 4. an enumeration of unresolved issues associated with the use of AI for control, and 5. a discussion of the possible role of expert system techniques in process control. (orig./GL)

  8. Analysis of man-machine interaction for control and display system in main control room of light water reactor

    International Nuclear Information System (INIS)

    Santosa, Kussigit; Supriatna, Piping; Karlina, Itjeu; Widagdo, Suharyo; Darlis; Sudiono, Bambang

    1998-01-01

    One of potential hazard in Nuclear Power Plant is the failure of its operation. The accident or operation failure in the reactor must be concerned event its probability is low. The important thing should be concerned is 'Analysis of Man-Machine Interaction (MMI) for Control and Display System in Main Control Room (MCR) of Nuclear Power Reactor', especially LWR type. Control and Display System in MCR of Reactor is the main part of MMI link process in Reactor MCR work system. Signal from display system showed performance process in reactor, while this signal will be received by operator. This signal will be described through central nerve for making decision what kind must be done. Then the operator manage the next process of reactor operation through control system. So by knowing Analysis of Man-Machine Interaction for Control and Display System in Main Control Room of Power Reactor, we can understand human error probability of the operator in reactor operation

  9. Control of reactor coolant flow path during reactor decay heat removal

    International Nuclear Information System (INIS)

    Hunsbedt, A.N.

    1988-01-01

    This patent describes a sodium cooled reactor of the type having a reactor hot pool, a slightly lower pressure reactor cold pool and a reactor vessel liner defining a reactor vessel liner flow gap separating the hot pool and the cold pool along the reactor vessel sidewalls and wherein the normal sodium circuit in the reactor includes main sodium reactor coolant pumps having a suction on the lower pressure sodium cold pool and an outlet to a reactor core; the reactor core for heating the sodium and discharging the sodium to the reactor hot pool; a heat exchanger for receiving sodium from the hot pool, and removing heat from the sodium and discharging the sodium to the lower pressure cold pool; the improvement across the reactor vessel liner comprising: a jet pump having a venturi installed across the reactor vessel liner, the jet pump having a lower inlet from the reactor vessel cold pool across the reactor vessel liner and an upper outlet to the reactor vessel hot pool

  10. Control Rod Driveline Reactivity Feedback Model for Liquid Metal Reactors

    International Nuclear Information System (INIS)

    Kwon, Young-Min; Jeong, Hae-Yong; Chang, Won-Pyo; Cho, Chung-Ho; Lee, Yong-Bum

    2008-01-01

    The thermal expansion of the control rod drivelines (CRDL) is one important passive mitigator under all unprotected accident conditions in the metal and oxide cores. When the CRDL are washed by hot sodium in the coolant outlet plenum, the CRDL thermally expands and causes the control rods to be inserted further down into the active core region, providing a negative reactivity feedback. Since the control rods are attached to the top of the vessel head and the core attaches to the bottom of the reactor vessel (RV), the expansion of the vessel wall as it heats will either lower the core or raise the control rods supports. This contrary thermal expansion of the reactor vessel wall pulls the control rods out of the core somewhat, providing a positive reactivity feedback. However this is not a safety factor early in a transient because its time constant is relatively large. The total elongated length is calculated by subtracting the vessel expansion from the CRDL expansion to determine the net control rod expansion into the core. The system-wide safety analysis code SSC-K includes the CRDL/RV reactivity feedback model in which control rod and vessel expansions are calculated using single-nod temperatures for the vessel and CRDL masses. The KALIMER design has the upper internal structures (UIS) in which the CRDLs are positioned outside the structure where they are exposed to the mixed sodium temperature exiting the core. A new method to determine the CRDL expansion is suggested. Two dimensional hot pool thermal hydraulic model (HP2D) originally developed for the analysis of the stratification phenomena in the hot pool is utilized for a detailed heat transfer between the CRDL mass and the hot pool coolant. However, the reactor vessel wall temperature is still calculated by a simple lumped model

  11. Digital Sliding Mode Control of Anti-Lock Braking System

    Directory of Open Access Journals (Sweden)

    MITIC, D. B.

    2013-02-01

    Full Text Available The control of anti-lock braking system is a great challenge, because of the nonlinear and complex characteristics of braking dynamics, unknown parameters of vehicle environment and system parameter variations. Using some of robust control methods, such as sliding mode control, can be a right solution for these problems. In this paper, we introduce a novel approach to design of ABS controllers, which is based on digital sliding mode control with only input/output measurements. The relay term of the proposed digital sliding mode control is filtered through digital integrator, reducing the chattering phenomenon in that way, and the additional signal of estimated modelling error is introduced into control algorithm to enhance the system steady-state accuracy. The given solution was verified in real experimental framework and the obtained results were compared with the results of implementation of two other digital sliding mode control algorithms. It is shown that it gives better system response, higher steady-state accuracy and smaller chattering.

  12. Use of hafnium in control bars of nuclear reactors; Uso de hafnio en barras de control de reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J.R.; Alonso V, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: jrrs@nuclear.inin-mx

    2003-07-01

    Recently the use of hafnium as neutron absorber material in nuclear reactors has been reason of investigation by virtue of that this material has nuclear properties as to the neutrons absorption and structural that can prolong the useful life of the control mechanisms of the nuclear reactors. In this work some of those more significant hafnium properties are presented like nuclear material. Also there are presented calculations carried out with the HELIOS code for fuel cells of uranium oxide and of uranium and plutonium mixed oxides under controlled conditions with conventional bars of boron carbide and also with similar bars to which are substituted the absorbent material by metallic hafnium, the results are presented in this work. (Author)

  13. Water quality control device and water quality control method for reactor primary coolant system

    International Nuclear Information System (INIS)

    Wada, Yoichi; Ibe, Eishi; Watanabe, Atsushi.

    1995-01-01

    The present invention is suitable for preventing defects due to corrosion of structural materials in a primary coolant system of a BWR type reactor. Namely, a concentration measuring means measures the concentration of oxidative ingredients contained in a reactor water. A reducing electrode is disposed along a reactor water flow channel in the primary coolant system and reduces the oxidative ingredients. A reducing counter electrode is disposed along the reactor water flow channel in the primary coolant system, and electrically connected to the reducing electrode. The reactor structural materials are used as a reference electrode providing a reference potential to the reducing electrode and the reducing counter electrode. A potential control means controls the potential of the reducing electrode relative to the reference potential based on the signals from the concentration measuring means. A stable reference potential in a region where an effective oxygen concentration is stable can be obtained irrespective of the change of operation conditions by using the reactor structural materials disposed to a boiling region in the reactor core as a reference electrode. As a result, the water quality can be controlled at high accuracy. (I.S.)

  14. A study of digital hardware architectures for nuclear reactors protection systems applications - reliability and safety analysis methods

    International Nuclear Information System (INIS)

    Benko, Pedro Luiz

    1997-01-01

    A study of digital hardware architectures, including experience in many countries, topologies and solutions to interface circuits for protection systems of nuclear reactors is presented. Methods for developing digital systems architectures based on fault tolerant and safety requirements is proposed. Directives for assessing such conditions are suggested. Techniques and the most common tools employed in reliability, safety evaluation and modeling of hardware architectures is also presented. Markov chain modeling is used to evaluate the reliability of redundant architectures. In order to estimate software quality, several mechanisms to be used in design, specification, and validation and verification (V and V) procedures are suggested. A digital protection system architecture has been analyzed as a case study. (author)

  15. A Digital Self Excited Loop for Accelerating Cavity Field Control

    International Nuclear Information System (INIS)

    Curt Hovater; Trent Allison; Jean Delayen; John Musson; Tomasz Plawski

    2007-01-01

    We have developed a digital process that emulates an analog oscillator and ultimately a self excited loop (SEL) for field control. The SEL, in its analog form, has been used for many years for accelerating cavity field control. In essence the SEL uses the cavity as a resonant circuit -- much like a resonant (tank) circuit is used to build an oscillator. An oscillating resonant circuit can be forced to oscillate at different, but close, frequencies to resonance by applying a phase shift in the feedback path. This allows the circuit to be phased-locked to a master reference, which is crucial for multiple cavity accelerators. For phase and amplitude control the SEL must be forced to the master reference frequency, and feedback provided for in both dimensions. The novelty of this design is in the way digital signal processing (DSP) is structured to emulate an analog system. While the digital signal processing elements are not new, to our knowledge this is the first time that the digital SEL concept has been designed and demonstrated. This paper reports on the progress of the design and implementation of the digital SEL for field control of superconducting accelerating cavities

  16. Advanced Instrumentation, Information, and Control Systems Technologies Research in Support of Light Water Reactors

    International Nuclear Information System (INIS)

    Hallbert, Bruce P.; Kenneth, Thomas

    2014-01-01

    The Advanced Instrumentation, Information, and Control (II and C) Systems Technologies Pathway conducts targeted research and development (R and D) to address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals to ensure that legacy analog II and C systems are not life-limiting issues for the LWR fleet, and to implement digital II and C technology in a manner that enables broad innovation and business improvement in the nuclear power plant operating model. Resolving long-term operational concerns with the II and C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation's energy and environmental security

  17. Advanced Instrumentation, Information, and Control Systems Technologies Research in Support of Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hallbert, Bruce P.; Kenneth, Thomas [Idaho National Laboratory, Idaho (United States)

    2014-08-15

    The Advanced Instrumentation, Information, and Control (II and C) Systems Technologies Pathway conducts targeted research and development (R and D) to address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals to ensure that legacy analog II and C systems are not life-limiting issues for the LWR fleet, and to implement digital II and C technology in a manner that enables broad innovation and business improvement in the nuclear power plant operating model. Resolving long-term operational concerns with the II and C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation's energy and environmental security.

  18. Digital repetitive control under varying frequency conditions

    CERN Document Server

    Ramos, Germán A; Olm, Josep M

    2013-01-01

    The tracking/rejection of periodic signals constitutes a wide field of research in the control theory and applications area. Repetitive Control has proven to be an efficient way to face this topic. However, in some applications the frequency of the reference/disturbance signal is time-varying or uncertain. This causes an important performance degradation in the standard Repetitive Control scheme. This book presents some solutions to apply Repetitive Control in varying frequency conditions without loosing steady-state performance. It also includes a complete theoretical development and experimental results in two representative systems. The presented solutions are organized in two complementary branches: varying sampling period Repetitive Control and High Order Repetitive Control. The first approach allows dealing with large range frequency variations while the second allows dealing with small range frequency variations. The book also presents applications of the described techniques to a Roto-magnet plant and...

  19. Calculation method for control rod dropping time in reactor

    International Nuclear Information System (INIS)

    Nogami, Takeki; Kato, Yoshifumi; Ishino, Jun-ichi; Doi, Isamu.

    1996-01-01

    If a control rod starts dropping, the dropping speed is rapidly increased, then settled substantially constant, rapidly decreased when it reaches a dash pot. A second detection signal generated by removing an AC component from a first detection signal is differentiated twice. The time when the maximum value among the twice differentiated values is generated is determined as a time when the control rods starts dropping. The time when minimum value among the twice differentiated values is generated is determined as a time when the control rod reaches the dash pot of the reactor. The measuring time within a range from the time when the control rod starts dropping to the time when the control rod reaches the dash pot of the reactor is determined. As a result, processing for the calculation of the dropping start time and dash pot reaching time of the control rod can be automatized. Further, it is suffice to conduct differentiation twice till the reaching time, which can facilitate the processing thereby enabling to determine a reliable time range. (N.H.)

  20. Electroremediation of air pollution control residues in a continuous reactor

    DEFF Research Database (Denmark)

    Jensen, Pernille Erland; Ferreira, Célia M. D.; Hansen, Henrik K.

    2010-01-01

    Air pollution control (APC) residue from municipal solid waste incineration is considered hazardous waste due to its alkalinity and high content of salts and mobile heavy metals. Various solutions for the handling of APC-residue exist, however most commercial solutions involve landfilling. A demand...... for environmental sustainable alternatives exists and electrodialysis could be such an alternative. The potential of electrodialysis for treating APC-residue is explored in this work by designing and testing a continuous-flow bench-scale reactor that can work with a high solids content feed solution. Experiments...... were made with raw residue, water-washed residue, acid washed residue and acid-treated residue with emphasis on reduction of heavy metal mobility. Main results indicate that the reactor successfully removes toxic elements lead, copper, cadmium and zinc from the feed stream, suggesting...

  1. Device for controlling a recirculation flow in a reactor

    International Nuclear Information System (INIS)

    Shida, Toichi; Tohei, Kazushige; Hirose, Masao; Nakamura, Hideo.

    1976-01-01

    Object: To provide an emergency cut-off valve in a recirculation system in a reactor to control the recirculation at the time of turbine trip or load cut-off, thereby relieving excessive increase in heat output of fuel. Structure: A recirculation pump is driven through a recirculation pump motor by an AC generator, which is driven by a driving motor through a fluid coupling, so that reactor water passes the emergency cut-off valve and recirculation flow stop valve and then passes a jet pump into the core. At the time of turbine trip or load cut-off, the emergency cut-off valve is closed by a hydraulic circuit, whereby core flow is merely decreased by 20 to 30% in a short period of time to restrain excessive increase in heat output. (Yoshino, Y.)

  2. Lessons learned in process control at the Halden Reactor Project

    International Nuclear Information System (INIS)

    Kennedy, W.G.

    1989-12-01

    This report provides a list of those findings particularly relevant to regulatory authorities that can be derived from the research and development activities in computerized process control conducted at the Halden Reactor Project. The report was prepared by a staff member of the US Nuclear Regulatory Commission working at Halden. It identifies those results that may be of use to regulatory organizations in three main areas: as support for new requirements, as part of regulatory evaluations of the acceptability of new methods and techniques, and in exploratory research and development of new approaches to improve operator performance. More than 200 findings arranged in nine major categories are presented. The findings were culled from Halden Reactor Project documents, which are listed in the report

  3. Design of a digital system for operational parameters simulation of IPR-R1 TRIGA nuclear research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lage, Aldo M.F.; Mesquita, Amir Z.; Felippe, Adriano de A.M., E-mail: aldo@cdtn.br, E-mail: amir@cdtn.br, E-mail: adrianoamfelippe@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN /CNEN-MG), Belo Horizonte, MG (Brazil); Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)

    2017-11-01

    The instrumentation of nuclear reactors is designed based on the reliability, redundancy and diversification of control systems. The monitoring of the parameters is of crucial importance with regard to the operational efficiency and safety of the installation. Since the first criticality of a nuclear reactor, achieved by Fermi et al. in 1942, there has been concern about the reliable monitoring of the parameters involved in the chain reaction. This paper presents the current stage of the system of simulation, which is under development at the CDTN, which intends to simulate the operation of the TRIGA IPR-R1 nuclear reactor, involving the evolution of neutron flux and reactor power related events. The system will be developed using LabVIEW® software, using the modern concept of virtual instruments (VIs) that are visualized in a video monitor. For the implementation of this model, computational tools and systems analysis are necessary, which help and facilitate the implementation of the simulator. In this article we will show some of these techniques and the initial design of the model to be implemented. The design of a computational system is of great importance, since it guides in the implementation stages and generates the documentation for later maintenance and updating of the computational system. It is noteworthy that the innovations developed in research reactors are normally used in power reactors. The relatively low costs enable research reactors to be an excellent laboratory for developing techniques for future reactors. (author)

  4. Design of a digital system for operational parameters simulation of IPR-R1 TRIGA nuclear research reactor

    International Nuclear Information System (INIS)

    Lage, Aldo M.F.; Mesquita, Amir Z.; Felippe, Adriano de A.M.

    2017-01-01

    The instrumentation of nuclear reactors is designed based on the reliability, redundancy and diversification of control systems. The monitoring of the parameters is of crucial importance with regard to the operational efficiency and safety of the installation. Since the first criticality of a nuclear reactor, achieved by Fermi et al. in 1942, there has been concern about the reliable monitoring of the parameters involved in the chain reaction. This paper presents the current stage of the system of simulation, which is under development at the CDTN, which intends to simulate the operation of the TRIGA IPR-R1 nuclear reactor, involving the evolution of neutron flux and reactor power related events. The system will be developed using LabVIEW® software, using the modern concept of virtual instruments (VIs) that are visualized in a video monitor. For the implementation of this model, computational tools and systems analysis are necessary, which help and facilitate the implementation of the simulator. In this article we will show some of these techniques and the initial design of the model to be implemented. The design of a computational system is of great importance, since it guides in the implementation stages and generates the documentation for later maintenance and updating of the computational system. It is noteworthy that the innovations developed in research reactors are normally used in power reactors. The relatively low costs enable research reactors to be an excellent laboratory for developing techniques for future reactors. (author)

  5. Study on Reactor Power Transient Characteristics (Reactor Training Experiments) –Control Rod Reactivity Calibration by Positive Period Method and other Experiment–

    OpenAIRE

    尾崎, 禎彦; Ozaki, Yoshihiko; 砂川, 武義; Sunagawa, Takeyoshi

    2014-01-01

    In this paper, it is reported about some experiments that have been carried out in the reactor training that targets sophomore of the department of applied nuclear engineering, FUT. Reactor of Kinki University Atomic Energy Research Institute (UTR-KINKI) was used for reactor training. When each critical state was achieved at different reactor output respectively in reactor operating, it was confirmed that the control rod position at that time does not change. Further, control rod reactivity c...

  6. Stress testing of digital flight-control system software

    Science.gov (United States)

    Rajan, N.; Defeo, P. V.; Saito, J.

    1983-01-01

    A technique for dynamically testing digital flight-control system software on a module-by-module basis is described. Each test module is repetitively executed faster than real-time with an exhaustive input sequence. Outputs of the test module are compared with outputs generated by an alternate, simpler implementation for the same input data. Discrepancies between the two sets of output indicate the possible presence of a software error. The results of an implementation of this technique in the Digital Flight-Control System Software Verification Laboratory are discussed.

  7. Inteligent control system for a CANDU 600 type reactor process

    International Nuclear Information System (INIS)

    Venescu, B.; Zevedei, D.; Jurian, M.; Venescu, R.

    2013-01-01

    The present paper is set on presenting a highly intelligent configuration, capable of controlling, without the need of the human factor, a complete nuclear power plant type of system, giving it the status of an autonomous system. The urge for such a controlling system is justified by the amount of drawbacks that appear in real life as disadvantages, loses and sometimes even inefficiency in the current controlling and comanding systems of the nuclear reactors. The application stands in the comand sent from the auxiliary feedwater flow control valves to the steam generators. As an environment fit for development I chose Matlab Simulink to simulate the behaviour of the process and the adjusted system. Comparing the results obtained after the fuzzy regulation with those obtained after the classical regulation, we can demonstrate the necessity of implementing artificial intelligence techniques in nuclear power plants and we can agree to the advantages of being able to control everything automatically. (authors)

  8. Research on reactor power controller based on artificial immune P and PID cascade control technology

    International Nuclear Information System (INIS)

    Cheng Shouyu; Peng Minjun; Liu Xinkai

    2014-01-01

    The Reactor Power control system usually adopts the traditional PID controller, the traditional PID controller can meet the operating requirements, but the control effect is not very good. In order to improve this condition, the paper proposes an immune P and PID cascade controller which based the immune mechanism of B-cell co-operating with T-cell, the nuclear power controller based on artificial immune is less reported. In order to verify and validate the control strategy, the designed controller debugs with the full-scope real-time simulation system of nuclear power plants. The simulation results shows that the immune controller can effectively improve the dynamic operating characteristics of the reactor system, and the immune controller is superior to the traditional PID controller in control performance. (authors)

  9. Heating control device for nuclear reactor cooling systems

    International Nuclear Information System (INIS)

    Kishigawa, Osamu.

    1981-01-01

    Purpose: To obtain a heating control device capable of surely preventing deformations and fluctures due to thermal stresses in equipments such as tanks and pipeways upon starting of reactor operation. Constitution: A heating control device for nuclear reactor cooling systems using metal coolants comprises a plurality of heaters disposed at each of the sections in the cooling systems for heating the coolants, temperature detectors for the detection of temperature at each of the sections in the cooling systems to be heated by the heaters, a circuit for judging the range of filling process of the metal coolants in the cooling systems, a coefficient change circuit for heating control for changing and setting the coefficient for the heating control by the heaters based on the information for the range judged by the judging circuit, and a circuit for controlling the input to each of the heaters based on the output signals for the coefficient change circuit and the signals from the temperature detectors. (Seki, T.)

  10. The task of control digital image compression

    OpenAIRE

    TASHMANOV E.B.; МАМАTOV М.S.

    2014-01-01

    In this paper we consider the relationship of control tasks and image compression losses. The main idea of this approach is to allocate structural lines simplified image and further compress the selected data

  11. Attitude Control of a Satellite by using Digital Signal Processing

    Directory of Open Access Journals (Sweden)

    Adirelle C. Santana

    2012-03-01

    Full Text Available This article has discussed the development of a three-axis attitude digital controller for an artificial satellite using a digital signal processor. The main motivation of this study is the attitude control system of the satellite Multi-Mission Platform, developed by the Brazilian National Institute for Space Research for application in different sort of missions. The controller design was based on the theory of the Linear Quadratic Gaussian Regulator, synthesized from the linearized model of the motion of the satellite, i.e., the kinematics and dynamics of attitude. The attitude actuators considered in this study are pairs of cold gas jets powered by a pulse width/pulse frequency modulator. In the first stage of the project development, a system controller for continuous time was studied with the aim of testing the adequacy of the adopted control. The next steps had included an analysis of discretization techniques, the setting time of sampling rate, and the testing of the digital version of the Linear Quadratic Gaussian Regulator controller in the MATLAB/SIMULINK. To fulfill the study, the controller was implemented in a digital signal processor, specifically the Blackfin BF537 from Analog Devices, along with the pulse width/pulse frequency modulator. The validation tests used a scheme of co-simulation, where the model of the satellite was simulated in MATLAB/SIMULINK, while the controller and modulator were processed in the digital signal processor with a tool called Processor-In-the-Loop, which acted as a data communication link between both environments.function and required time to achieve a given mission accuracy are determined, and results are provided as illustration.

  12. Digital Speed Cascade Control, using Scilab / Xcos Environment

    Directory of Open Access Journals (Sweden)

    Petru Chioncel

    2016-10-01

    Full Text Available This paper presents an application of digital cascade control implemented in Scilab / Xcos environment, using a P type regulator for the position adjustment circuit, a PI controller for the speed circuit adjustment; the current respectively moments control circuits are rendered by elements of PT1 type. On this basis the program is done in Scilab and the related signal block diagram implemented in Xcos; through simulation, the step response of the system is analyzed for different sampling times.

  13. Digital Information Platform Design of Fuel Element Engineering For High Temperature Gas-cooled Reactor

    International Nuclear Information System (INIS)

    Du Yuwei

    2014-01-01

    This product line provide fuel element for high temperature gas-cooled reactor nuclear power plant which is being constructed in Shidao bay in Shandong province. Its annual productive capacity is thirty ten thousands fuel elements whose shape is spherical . Compared with pressurized water fuel , this line has the feature of high radiation .In order to reduce harm to operators, the comprehensive information platform is designed , which can realize integration of automation and management for plant. This platform include two nets, automation net using field bus technique and information net using Ethernet technique ,which realize collection ,control, storage and publish of information.By means of construction, automatization and informatization of product line can reach high level. (author)

  14. Development and testing of a recorder and controller for a microalgae culture reactor; Desarrollo y prueba de un registrador y controlador para un reactor de cultivo de microalgas

    Energy Technology Data Exchange (ETDEWEB)

    Esquivel, Wilson; Reyes, Jose Fernando; Bruijn, Johannes; Hernandez, Alejandro [Universidad de Concepcion, Chilan (Chile). Facultad de Ingenieria Agricola. Dept. de Mecanizacion y Energia], Emails: wesquive@udec.cl., jreyes@udec.cl., jdebruij@udec.cl., alehernandez@udec.cl

    2010-07-01

    An electronic system to monitor and control operational variables in a Raceway type of reactor for the culture of the Scenedesmus spinosus microalgae and later production for biodiesel and mitigating CO{sub 2} was developed and tested. The electronic system is constituted by a micro controller, a card reader SD, a card SD, a real-time clock, a power supply, a screen GLCD, a keyboard and a card for data acquisition, all implemented for 4-20 mA and 0-5 V output sensors. Temperature, pH, electrical conductivity, dissolved oxygen and solar radiation were measured digitalized and saved every 10 minutes. These variables were digitalized and kept in the SD memory every 10 minutes. It was determined that the most favorable conditions for the proliferation of the culture are near pH neutral and a temperature of 30 deg C, existing a strong correlation between pH and the dissolved CO{sub 2} level. Using the digital outputs of temperature and pH of the microcontroller, the CO{sub 2} injection and the elimination of O{sub 2} were controlled to maintain an adequate environment for the development of the culture. (author)

  15. Plasma control issues for an advanced steady state tokamak reactor

    International Nuclear Information System (INIS)

    Moreau, D.

    2001-01-01

    This paper deals with specific control issues related to the advanced tokamak scenarios in which rather accurate tailoring of the current density profile is a requirement in connection with the steady state operation of a reactor in a high confinement optimized shear mode. It is found that adequate current profile control can be performed if real-time magnetic flux reconstruction is available through a set of dedicated diagnostics and computers, with sufficient accuracy to deduce the radial profile of the safety factor and of the internal plasma loop voltage. It is also shown that the safety factor can be precisely controlled in the outer half of the plasma through the surface loop voltage and the off-axis current drive power, but that a compromise must be made between the accuracy of the core safety factor control and the total duration of the current and fuel density ramp-up phases, so that the demonstration of the steady state reactor potential of the optimized/reversed shear concept in the Next Step device will demand pulse lengths of the order of one thousand seconds (or more for an ITER-size machine). (author)

  16. Improvements to the TRIGA Mark II instrumentation and the direct digital control by microprocessors

    International Nuclear Information System (INIS)

    Tomsic, M.; Taves, R.; Mrcun, I.; Kavsek, D.; Znidaric, B.

    1978-01-01

    Two tendencies have been present in the maintenance of the TRIGA instrumentation: one was to renew only those parts that were deteriorating with age, thus ensuring the continuation of the satisfactory service in the original scope; the other was aimed at adding new features and possibly at changing the whole concept of the reactor control and instrumentation. Though the activities along both lines were not best coordinated at all times, the presently emerging result may be highly satisfactory. Besides the well maintained instrumentation in the original scope and concept, a digital data logging and control system is being installed, based on microprocessors, which should offer new level of flexibility and convenience to the operators and experimenters, without compromising either safety of reliability of the overall instrumentation and control system

  17. Computation of reactor control rod drop time under accident conditions

    International Nuclear Information System (INIS)

    Dou Yikang; Yao Weida; Yang Renan; Jiang Nanyan

    1998-01-01

    The computational method of reactor control rod drop time under accident conditions lies mainly in establishing forced vibration equations for the components under action of outside forces on control rod driven line and motion equation for the control rod moving in vertical direction. The above two kinds of equations are connected by considering the impact effects between control rod and its outside components. Finite difference method is adopted to make discretization of the vibration equations and Wilson-θ method is applied to deal with the time history problem. The non-linearity caused by impact is iteratively treated with modified Newton method. Some experimental results are used to validate the validity and reliability of the computational method. Theoretical and experimental testing problems show that the computer program based on the computational method is applicable and reliable. The program can act as an effective tool of design by analysis and safety analysis for the relevant components

  18. Drive reinforcement neural networks for reactor control. Final report

    International Nuclear Information System (INIS)

    Williams, J.G.; Jouse, W.C.

    1995-01-01

    In view of the loss of the third year funding, the scope of the project goals has been revised. The revision in project scope no longer allows for the detailed modeling of the EBR-11 start-up task that was originally envisaged. The authors are continuing, however, to model the control of the rapid power ascent of the University of Arizona TRIGA reactor using a model-based controller and using a drive reinforcement neural network. These will be combined during the concluding period of the project into a hierarchical control architecture. In addition, the modeling of a PWR feedwater heater has continued, and an autonomous fault-tolerant software architecture for its control has been proposed

  19. Robust reactor power control system design by genetic algorithm

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yoon Joon; Cho, Kyung Ho; Kim, Sin [Cheju National University, Cheju (Korea, Republic of)

    1997-12-31

    The H{sub {infinity}} robust controller for the reactor power control system is designed by use of the mixed weight sensitivity. The system is configured into the typical two-port model with which the weight functions are augmented. Since the solution depends on the weighting functions and the problem is of nonconvex, the genetic algorithm is used to determine the weighting functions. The cost function applied in the genetic algorithm permits the direct control of the power tracking performances. In addition, the actual operating constraints such as rod velocity and acceleration can be treated as design parameters. Compared with the conventional approach, the controller designed by the genetic algorithm results in the better performances with the realistic constraints. Also, it is found that the genetic algorithm could be used as an effective tool in the robust design. 4 refs., 6 figs. (Author)

  20. Applying human factors to the design of control centre and workstation of a nuclear reactor

    International Nuclear Information System (INIS)

    Santos, Isaac J.A. Luquetti dos; Carvalho, Paulo V.R.; Goncalves, Gabriel de L.; Souza, Tamara D.M.F.; Falcao, Mariana A.

    2013-01-01

    Human factors is a body of scientific factors about human characteristics, covering biomedical, psychological and psychosocial considerations, including principles and applications in the personnel selection areas, training, job performance aid tools and human performance evaluation. Control Centre is a combination of control rooms, control suites and local control stations which are functionally related and all on the same site. Digital control room includes an arrangement of systems, equipment such as computers and communication terminals and workstations at which control and monitoring functions are conducted by operators. Inadequate integration between control room and operators reduces safety, increases the operation complexity, complicates operator training and increases the likelihood of human errors occurrence. The objective of this paper is to present a specific approach for the conceptual and basic design of the control centre and workstation of a nuclear reactor used to produce radioisotope. The approach is based on human factors standards, guidelines and the participation of a multidisciplinary team in the conceptual and basic phases of the design. Using the information gathered from standards and from the multidisciplinary team, an initial sketch 3D of the control centre and workstation are being developed. (author)

  1. Digital feed back control for radial beam position

    International Nuclear Information System (INIS)

    Mestha, L.K.

    1989-09-01

    In the development of wide spread large scale distributed digital control systems, there is a requirement to automate small processes like radial beam control which will not only improve the beam quality but will also add local intelligence. Hence use is made here of digital control principles for such applications. The work concerned with the radial beam control discussed in this report has been developed for ISIS at RAL. The structure of the report is hence inclined more towards the local hardware system. The general feed back loop techniques can also be implemented for other control purpose. For instance, the author has successfully tested similar techniques to minimise the RF cavity tuning error, where the improvement in performance could not be matched by the analogue loop. A description of the RF cavity tuning programme and the associated experimental results will be published as a local paper for ISIS division. (author)

  2. Assessment of Digital Access Control Methods Used by Selected ...

    African Journals Online (AJOL)

    Assessment of Digital Access Control Methods Used by Selected Academic Libraries in South-West Nigeria. ... information professionals with the knowledge that would enable them establish an effective strategy to protect e-resources from such abuses as plagiarism, piracy and infringement of intellectual property rights.

  3. digital control of external devices through the parallel port

    African Journals Online (AJOL)

    2012-11-03

    Nov 3, 2012 ... Abstract. In this paper we carry out the digital control of external devices using the parallel port of a computer. The PC parallel port adapter that is specifically designed to attach printers has been found to be useful as a general input/output port for any device or application that matches its input/output ...

  4. Issues and approaches in control for autonomous reactor operation

    International Nuclear Information System (INIS)

    Vilim, R. B.; Khalil, H. S.; Wei, T. Y. C.

    2000-01-01

    A capability for autonomous and passively safe operation is one of the goals of the NERI funded development of Generation IV nuclear plants. An approach is described for evaluating the effect of increasing autonomy on safety margins and load behavior and for examining issues that arise with increasing autonomy and their potential impact on performance. The method provides a formal approach to the process of exploiting the innate self-regulating property of a reactor to make it less dependent on operator action and less vulnerable to automatic control system fault and/or operator error. Some preliminary results are given

  5. Clinch River Breeder Reactor secondary control rod system

    International Nuclear Information System (INIS)

    McKeehan, E.R.; Sim, R.G.

    1977-01-01

    The shutdown system for the Clinch River Breeder Reactor (CRBR) includes two independent systems--a primary and a secondary system. The Secondary Control Rod System (SCRS) is a new design which is being developed by General Electric to be independent from the primary system in order to improve overall shutdown reliability by eliminating potential common-mode failures. The paper describes the status of the SCRS design and fabrication and testing activities. Design verification testing on the component level is largely complete. These component tests are covered with emphasis on design impact results. A prototype unit has been manufactured and system level tests in sodium have been initiated

  6. Fragile imperceptible digital watermark with privacy control

    Science.gov (United States)

    Coppersmith, Don; Mintzer, Frederick C.; Tresser, Charles P.; Wu, Chai W.; Yeung, Minerva M.

    1999-04-01

    We propose a watermarking scheme which allows the watermarked image to be authenticated by an authentication agent without revealing to the authentication agent the human-readable content of the image by combining privacy control with watermarking and authentication mechanisms. This watermarking scheme has universal applicability to data sets such as image, video and audio bit streams. The watermark can be made to be imperceptible to humans. Usage of public key cryptography allows the authentication agent to authenticate without the capabilities to watermark an image.

  7. A high-performance digital control system for TCV

    International Nuclear Information System (INIS)

    Lister, J.B.; Dutch, M.J.; Milne, P.G.; Means, R.W.

    1997-10-01

    The TCV hybrid analogue-digital plasma control system has been superseded by a high performance Digital Plasma Control System, DPCS, made possible by recent advances in off the shelf technology. We discuss the basic requirements for such a control system and present the design and specifications which were laid down. The nominal and final performances are presented and the complete design is given in detail. The integration of the new system into the current operation of the TCV tokamak is described. The procurement of this system has required close collaboration between the end-users and two commercial suppliers with one of the latter taking full responsibility for the system integration. The impact of this approach on the design and commissioning costs for the TCV project is presented. New possibilities offered by this new system are discussed, including possible work relevant to ITER plasma control development. (author) 3 figs., 5 refs

  8. Compactable control element assembly for a nuclear reactor

    International Nuclear Information System (INIS)

    Dupen, C.F.G.

    1976-01-01

    A description is given of a compactable control element assembly for a nuclear reactor in which the absorber pins of the assembly are compacted during downward movement of the pin and are returned to their uncompacted state when downward movement is stopped. The control element assembly comprises a support member longitudinally movable within a control assembly duct and a plurality of absorber pins supported laterally outward of the support member and within the duct by pairs of support arms. The absorber pins are pivotably mounted to the support arms and the support arms in turn are supported from the support member for upward pivotable movement in a longitudinal plane. As the support member is moved downward, the support arms pivot upwardly and the absorber pins move upwardly and inwardly towards the support member. When the support member is stopped the absorber pins return to their uncompacted position

  9. Capacitor requirements for controlled thermonuclear experiments and reactors

    International Nuclear Information System (INIS)

    Boicourt, G.P.; Hoffman, P.S.

    1975-01-01

    Future controlled thermonuclear experiments as well as controlled thermonuclear reactors will require substantial numbers of capacitors. The demands on these units are likely to be quite severe and quite different from the normal demands placed on either present energy storage capacitors or present power factor correction capacitors. It is unlikely that these two types will suffice for all necessary Controlled Thermonuclear Research (CTR) applications. The types of capacitors required for the various CTR operating conditions are enumerated. Factors that influence the life, cost and operating abilities of these types of capacitors are discussed. The problems of capacitors in a radiation environment are considered. Areas are defined where future research is needed. Some directions that this research should take are suggested. (U.S.)

  10. Application of controlled thermonuclear reactor fusion energy for food production

    International Nuclear Information System (INIS)

    Dang, V.D.; Steinberg, M.

    1975-06-01

    Food and energy shortages in many parts of the world in the past two years raise an immediate need for the evaluation of energy input in food production. The present paper investigates systematically (1) the energy requirement for food production, and (2) the provision of controlled thermonuclear fusion energy for major energy intensive sectors of food manufacturing. Among all the items of energy input to the ''food industry,'' fertilizers, water for irrigation, food processing industries, such as beet sugar refinery and dough making and single cell protein manufacturing, have been chosen for study in detail. A controlled thermonuclear power reactor was used to provide electrical and thermal energy for all these processes. Conceptual design of the application of controlled thermonuclear power, water and air for methanol and ammonia synthesis and single cell protein production is presented. Economic analysis shows that these processes can be competitive. (auth)

  11. 10 CFR 50.44 - Combustible gas control for nuclear power reactors.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 1 2010-01-01 2010-01-01 false Combustible gas control for nuclear power reactors. 50.44... for nuclear power reactors. (a) Definitions—(1) Inerted atmosphere means a containment atmosphere with... pressurized water nuclear power reactor with an operating license on October 16, 2003, except for those...

  12. Evaluation and quality control of digital subtraction angiography systems

    International Nuclear Information System (INIS)

    Louisot, P.

    1986-04-01

    After reviewing the development of systems used in angiography, we rewind the medical interest and describe the steps of an angiographic examination. The following chapter is dedicated to the techniques used for the digitalization of video images. The components of the system involved in the image acquisition are thoroughly investigated in chapter 4. Then, we analyse the capabilities of the machines available in France in 1985. Chapter 6 is devoted to the criteria of quality in digital imaging. In order to assign qualitative values to the above criteria, we design a control procedure which is described in chapter 7. The procedure thus allows the estimate of the physical performances of angiographic digital subtraction systems [fr

  13. Photovoltaic System Equipped with Digital Command Control and Acquisition

    OpenAIRE

    Yaden, Med; Melhaoui, Mustapha; Gaamouche, Rajae; Hirech, Kamal; Baghaz, Elhadi; Kassmi, Khalil

    2013-01-01

    In this paper, we present results concerning the design, the realization and the characterization of a photovoltaic system (PV), equipped with a digital controls: Power Point Tracking (MPPT), charge/discharge lead acid batteries, sun tracker and supervision. These different functions are performed with a microcontroller that has capabilities and functions to the reliability of PV systems (signal generation Pulses Width Modulation (PWM), speed etc.). Concerning the MPPT control operation, we i...

  14. An evaluation of control rod motion simulator of research reactor

    International Nuclear Information System (INIS)

    Sanda

    2010-01-01

    Motion simulator for rod control research reactor has been carried out using a servo motor. Reactor rod motion control at any point should be in the right position, one of the motors that can move in a precise and correct is the servo motor. To ensure that the servo motor to move in accordance with the desired program, then the servo motor function test should be carried out to ensure having good performance. Tests carried out on meshes stress disorder, the load is stable within a certain period and travel time safety control rod up and down, travel time regulating control rods up and down and travel time compensation control rods up and down. In testing the breakdown voltage V out nets at 24 V, 6.5 A with 12 Q load deviation obtained V0= V1 = 0.1% and 0.65% and for the stability of the load in a certain time deviation V = 0.7125% , next to the breakdown voltage V out nets at 12 V, 4.2 A with a 6 Q load deviation obtained V0= V1 = 0.275% and 1.158% for the stability of the load in a certain time deviation V = 1.463% and the net-voltage noise nets on V out 24 V, 4.5 A with 12 Q load deviation obtained V0 = V1 = 0.196% and 0.496% and for the stability of the load in a certain time deviation V = 0.3625%. While the travel time of a safety control rod up and down, up and down the regulator and compensation rise and fall showed a steady linear graph. The results show that the performance of the servo motor is very stable with the working area below the tolerance limit, it is 5% - 10%.(author)

  15. Nonlinear Control of Hydraulic Manipulator for Decommissioning Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myoung-Ho; Lee, Sung-Uk; Kim, Chang-Hoi; Choi, Byung-Seon; Moon, Jei-Kwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Robot technique is need to decommission nuclear reactor because of high radiation environment. Especially, Manipulator systems are useful for dismantling complex structure in a nuclear facility. In addition, Hydraulic system is applied to handle heavy duty object. Since hydraulic system can demonstrate high power. The manipulator with hydraulic power is already developed. To solve this problem, various nonlinear control method includes acceleration control. But, it is difficult because acceleration value is highly noisy. In this paper, the nonlinear control algorithm without acceleration control is studied. To verify, the hydraulic manipulator model had been developed. Furthermore, the numerical simulation is carried out. The nonlinear control without acceleration parameter method is developed for hydraulic manipulator. To verify control algorithm, the manipulator is modeled by MBD and the hydraulic servo system is also derived. In addition, the numerical simulation is also carried out. Especially, PID gain is determined though TDC algorithm. In the result of numerical simulation, tracking performance is good without acceleration control. Thus, the PID though TDC with SMC is good for hydraulic manipulator control.

  16. Self-tuning fuzzy logic nuclear reactor controller

    International Nuclear Information System (INIS)

    Sharif Heger, A.; Alang-Rashid, N.K.

    1996-01-01

    We present a method for self-tuning of fuzzy logic controllers based on the estimation of the optimum value of the centroids of its output fuzzy set. The method can be implemented on-line and does not require modification of membership functions and control rules. The main features of this method are: the rules are left intact to retain the operator's expertise in the FLC rule base, and the parameters that require any adjustment are identifiable in advance and their number is kept at a minimum. Therefore, the use of this method preserves the control statements in the original form. Results of simulation and actual tests show that this tuning method improves the performance of fuzzy logic controllers in following the desired reactor power level trajectories. In addition, this method demonstrates a similar improvement for power up and power down experiments, based on both simulation and actual case studies. For these experiments, the control rules for the fuzzy logic controller were derived from control statements that expressed the relationships between error, rate of error change, and duration of direction of control rod movements

  17. Instrumentation and control systems that can be used in a research reactor. Annex

    International Nuclear Information System (INIS)

    2015-01-01

    The instrumentation and control systems of a research reactor involve many systems that may differ depending on the type of reactor, the purpose and its modes of operation. Usually, it would include those systems identified in Section 2 as examples of instrumentation and control systems. Typical sets of instrumentation and control systems and their interrelations are shown. This Annex identifies instrumentation and control systems that can be used in a research reactor. Some of these instrumentation and control systems might not be used in a particular research reactor if they are not required for that specific type of installation

  18. Software verification and validation methodology for advanced digital reactor protection system using diverse dual processors to prevent common mode failure

    International Nuclear Information System (INIS)

    Son, Ki Chang; Shin, Hyun Kook; Lee, Nam Hoon; Baek, Seung Min; Kim, Hang Bae

    2001-01-01

    The Advanced Digital Reactor Protection System (ADRPS) with diverse dual processors is being developed by the National Research Lab of KOPEC for ADRPS development. One of the ADRPS goals is to develop digital Plant Protection System (PPS) free of Common Mode Failure (CMF). To prevent CMF, the principle of diversity is applied to both hardware design and software design. For the hardware diversity, two different types of CPUs are used for Bistable Processor and Local Coincidence Logic Processor. The VME based Single Board Computers (SBC) are used for the CPU hardware platforms. The QNX Operating System (OS) and the VxWorks OS are used for software diversity. Rigorous Software Verification and Validation (V and V) is also required to prevent CMF. In this paper, software V and V methodology for the ADRPS is described to enhance the ADRPS software reliability and to assure high quality of the ADRPS software

  19. Performance of static var compensator control type thyristor controlled reactor and thyristor switched capacitor

    Energy Technology Data Exchange (ETDEWEB)

    Araujo, Josias M. de; Yung, Chou Shaw; Rose, Eber H.; Pantoja, Antonio L.A. [ELETRONORTE, Belem, PA (Brazil); Fouesnant, Thomas; Boissier, Luc

    1994-12-31

    This paper has the objective of presenting the philosophy of Static Var Compensator (SVC) Control as well the necessary adjustments in the project of control system to guarantee suitable performance under different operating conditions. The verification on the performance of the SVC control has been done by Transient Network Analyzer (TNA/CEPEL) studies, commissioning tests and a factory tests. The SVC is the type of Thyristor Controlled Reactor (TCR) and Thyristor Switched Capacitor (TSC). (author) 3 refs., 12 figs.

  20. Development of automated controller system for controlling reactivity by using FPGA in research reactor application

    International Nuclear Information System (INIS)

    Mohd Sabri Minhat; Izhar Abu Hussin; Mohd Idris Taib

    2012-01-01

    The scope for this research paper is to produce a detail design for Development of Automated Controller System for Controlling Reactivity by using FPGA in Research Reactor Application for high safety nuclear operation. The development of this project including design, purchasing, fabrication, installation, testing and validation and verification for one prototype automated controller system for controlling reactivity in industry local technology for human capacity and capability development towards the first Nuclear Power Programme (NPP) in Malaysia. The specific objectives of this research paper are to Development of Automated Controller System for Controlling Reactivity (ACSCR) in Research Reactor Application (PUSPATI TRIGA Reactor) by using simultaneous movement method; To design, fabricate and produce the accuracy of Control Rods Drive Mechanism to 0.1 mm resolution using a stepper motor as an actuator; To design, install and produce the system response to be more faster by using Field Programmable Gate Array (FPGA) and High Speed Computer; and to improve the Safety Level of the Research Reactor in high safety nuclear operation condition. (author)

  1. Quality control in digital mammography: the noise components

    Energy Technology Data Exchange (ETDEWEB)

    Leyton, Fernando [Universidade de Tarapaca, Arica (Chile). Centro de Estudios en Ciencias Radiologicas; Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Nogueira, Maria do Socorro, E-mail: mnogue@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Duran, Maria Paz [Clinica Alemana, Santiago (Chile). Dept. de Radiologia; Dantas, Marcelino, E-mail: marcelino@inb.gov.b [Industrias Nucleares do Brasil (INB), Caldas, MG (Brazil). Unidade de Tratamento de Minerios; Ubeda, Carlos, E-mail: cubeda@uta.c [Universidade de Tarapaca, Arica (Chile). Fac. de Ciencias de la Salud

    2011-07-01

    To measure the linearity of the detector and determine the noise components (quantum, electronic and structural noise) that contributed to losing image quality and to determine the signal noise ratio (SNR) and contrast noise ratio (CNR). This paper describes the results of the implementation of a protocol for quality control in digital mammography performed in two direct digital mammography equipment (Hologic, Selenia) in Santiago of Chile. Shows the results of linearity and noise analysis of the images which establishes the main cause of noise in the image of the mammogram to ensure the quality and optimize procedures. The study evaluated two digital mammography's Selenia, Hologic (DR) from Santiago, Chile. We conducted the assessment of linearity of the detector, the signal noise ratio, contrast noise ratio and was determined the contribution of different noise components (quantum, electronics and structural noise). Used different thicknesses used in clinical practice according to the protocol for quality control in digital mammography of Spanish society of medical physics and NHSBSP Equipment Report 0604 Version 3. The Selenia mammography software was used for the analysis of images and Unfors Xi detector for measuring doses. The mammography detector has a linear performance, the CNR and SNR did not comply with the Protocol for the thicknesses of 60 and 70 mm. The main contribution of the noise corresponds to the quantum noise, therefore it is necessary to adjust and optimize the mammography system. (author)

  2. Quality control in digital mammography: the noise components

    International Nuclear Information System (INIS)

    Leyton, Fernando; Nogueira, Maria do Socorro; Duran, Maria Paz; Dantas, Marcelino; Ubeda, Carlos

    2011-01-01

    To measure the linearity of the detector and determine the noise components (quantum, electronic and structural noise) that contributed to losing image quality and to determine the signal noise ratio (SNR) and contrast noise ratio (CNR). This paper describes the results of the implementation of a protocol for quality control in digital mammography performed in two direct digital mammography equipment (Hologic, Selenia) in Santiago of Chile. Shows the results of linearity and noise analysis of the images which establishes the main cause of noise in the image of the mammogram to ensure the quality and optimize procedures. The study evaluated two digital mammography's Selenia, Hologic (DR) from Santiago, Chile. We conducted the assessment of linearity of the detector, the signal noise ratio, contrast noise ratio and was determined the contribution of different noise components (quantum, electronics and structural noise). Used different thicknesses used in clinical practice according to the protocol for quality control in digital mammography of Spanish society of medical physics and NHSBSP Equipment Report 0604 Version 3. The Selenia mammography software was used for the analysis of images and Unfors Xi detector for measuring doses. The mammography detector has a linear performance, the CNR and SNR did not comply with the Protocol for the thicknesses of 60 and 70 mm. The main contribution of the noise corresponds to the quantum noise, therefore it is necessary to adjust and optimize the mammography system. (author)

  3. Method of controlling power of a heavy water reactor

    International Nuclear Information System (INIS)

    Masuda, Hiroyuki.

    1975-01-01

    Object: To adjust a level of heavy water in a region of reflection body to control power in a heavy water reactor. Structure: The interior of a core tank filled with heavy water is divided by a partition into a core heavy water region and a reflection body region formed by surrounding the core heavy water region, and a level of heavy water within the reflection body region is adjusted to control power. Preferably, it is desirable to communicate the core heavy water region with the reflection body heavy water region at their lower portion, and gas pressure applied to an upper portion within at least one of said regions is adjusted to adjust the level of heavy water within the reflection body heavy water region. Thereby, the heavy water within the reflection body heavy water region may be introduced into the core region, thus requiring no tank which stores heavy water within the reflection body region. (Kamimura, M.)

  4. Operation control device for a nuclear reactor fuel exchanger

    International Nuclear Information System (INIS)

    Aida, Takashi.

    1984-01-01

    Purpose: To provide a operation control device for a nuclear reactor fuel exchanger with reduced size and weight capable of optionally meeting the complicated and versatile mode of the operation scope. Constitution: The operation range of a fuel exchanger is finely divided so as to attain the state capable of discriminating between operation-allowable range and operation-inhibitive range, which are stored in a memory circuit. Upon operating the fuel exchanger, the position is detected and a divided range data corresponding to the present position is taken out from the memory circuit so as to determine whether the fuel exchanger is to be run or stopped. Use of reduced size and compact IC circuits (calculation circuit, memory circuit, data latch circuit) and input/output interface circuits or the likes contributes to the size reduction of the exchanger control system to enlarge the floor maintenance space. (Moriyama, K.)

  5. Design and optimization of fuzzy-PID controller for the nuclear reactor power control

    International Nuclear Information System (INIS)

    Liu Cheng; Peng Jinfeng; Zhao Fuyu; Li Chong

    2009-01-01

    This paper introduces a fuzzy proportional-integral-derivative (fuzzy-PID) control strategy, and applies it to the nuclear reactor power control system. At the fuzzy-PID control strategy, the fuzzy logic controller (FLC) is exploited to extend the finite sets of PID gains to the possible combinations of PID gains in stable region and the genetic algorithm to improve the 'extending' precision through quadratic optimization for the membership function (MF) of the FLC. Thus the FLC tunes the gains of PID controller to adapt the model changing with the power. The fuzzy-PID has been designed and simulated to control the reactor power. The simulation results show the favorable performance of the fuzzy-PID controller.

  6. Computerized supervision and control system for movement at the RP-10 reactor control rods bank

    International Nuclear Information System (INIS)

    Padilla M, C.E.

    1998-01-01

    The project involves the use of a compatible microcomputer, Labwindows/CVI software, as well as National Instruments data acquisition cards AT-MIO16-E10 and PC-DIO96 to modify the sequence of movement of the reactor's rods and control them from a graphic interface in a computer's monitor. This graphic presentation is set as console of virtual instruments from where rod movement can be conducted. Normal rod movement, bank rod movement, and rod calibration have been considered. These experiences involve different logic of rod movements, which will determine movement sequence. Control of the automatic range of a current amplifier module was also considered. This module is know as 'automatic pilot amplifier' and given the strategic location of its detector (compensated ionizing camera) at the reactor's core, it delivers neutron flux current considered as reference to superficial neutron flux distribution at the reactor's core. Lecture and monitoring of this signal allows taking the reactor to a certain power, current of this signal is proportional to the power we want the reactor to reach. Advantages obtained with this system include the update of the control console, more uniform distribution of neutron flux, with lower and uniform burnup of nuclear fuel. (author)

  7. Control de temperatura de un reactor químico utilizando estrategias de control optimo y adaptivo.

    OpenAIRE

    Benites Saravia, Nicanor Raúl

    2011-01-01

    El presente trabajo de investigación trata sobre el diseño de estrategias de Control Optimo y Adaptivo para el control de temperatura de un reactor químico a escala reactor a escala de 0.37548342 m3. El reactor químico a considerar es un reactor exotérmico; es decir un reactor de tanque agitado continuo (CSTR), el cual consta de un tanque con agitación casi perfecta, en el que hay un flujo continuo de material reaccionante y desde el cual sale continuamente el material que ha reaccionado (mat...

  8. Apparatus for coupling and rotatably securing control rod systems of nuclear reactors

    International Nuclear Information System (INIS)

    Jabsen, F.S.

    1977-01-01

    An apparatus is described for coupling and rotatably securing the control rod system of a nuclear reactor having a control rod assembly, a fuel assembly and a drive to translate longitudinally the control rod assembly within the fuel assembly. This apparatus is of importance during reactor maintenance when fuel assemblies are removed from the core for refuelling and is particularly applicable in the case of the consolidated nuclear steam generator type of reactor. (U.K.)

  9. Evaluation of 'period-generated' control laws for the time-optimal control of reactor power

    International Nuclear Information System (INIS)

    Bernard, J.A.

    1988-01-01

    Time-Optimal control of neutronic power has recently been achieved by developing control laws that determine the actuator mechanism velocity necessary to produce a specified reactor period. These laws are designated as the 'MIT-SNL Period-Generated Minimum Time Control Laws'. Relative to time-optimal response, they function by altering the rate of change of reactivity so that the instantaneous period is stepped from infinity to its minimum allowed value, held at that value until the desired power level is attained, and then stepped back to infinity. The results of a systematic evaluation of these laws are presented. The behavior of each term in the control laws is shown and the capability of these laws to control properly the reactor power is demonstrated. Factors affecting the implementation of these laws, such as the prompt neutron lifetime and the differential reactivity worth of the actuators, are discussed. Finally, the results of an experimental study in which these laws were used to adjust the power of the 5 MWt MIT Research Reactor are shown. The information presented should be of interest to those designing high performance control systems for test, spacecraft, or, in certain instances, commercial reactors

  10. Supervisory Control System Architecture for Advanced Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cetiner, Sacit M [ORNL; Cole, Daniel L [University of Pittsburgh; Fugate, David L [ORNL; Kisner, Roger A [ORNL; Melin, Alexander M [ORNL; Muhlheim, Michael David [ORNL; Rao, Nageswara S [ORNL; Wood, Richard Thomas [ORNL

    2013-08-01

    This technical report was generated as a product of the Supervisory Control for Multi-Modular SMR Plants project within the Instrumentation, Control and Human-Machine Interface technology area under the Advanced Small Modular Reactor (SMR) Research and Development Program of the U.S. Department of Energy. The report documents the definition of strategies, functional elements, and the structural architecture of a supervisory control system for multi-modular advanced SMR (AdvSMR) plants. This research activity advances the state-of-the art by incorporating decision making into the supervisory control system architectural layers through the introduction of a tiered-plant system approach. The report provides a brief history of hierarchical functional architectures and the current state-of-the-art, describes a reference AdvSMR to show the dependencies between systems, presents a hierarchical structure for supervisory control, indicates the importance of understanding trip setpoints, applies a new theoretic approach for comparing architectures, identifies cyber security controls that should be addressed early in system design, and describes ongoing work to develop system requirements and hardware/software configurations.

  11. A Low-Power Digitally Controlled Oscillator for All Digital Phase-Locked Loops

    Directory of Open Access Journals (Sweden)

    Jun Zhao

    2010-01-01

    Full Text Available A low-power and low-jitter 12-bit CMOS digitally controlled oscillator (DCO design is presented. The Low-Power CMOS DCO is designed based on the ring oscillator implemented with Schmitt trigger inverters. The proposed DCO circuit uses control codes of thermometer type to reduce jitters. Performance of the DCO is verified through a novel All Digital Phase-Locked Loop (ADPLL designed with a unique lock-in process by employing a time-to-digital converter, where both the frequency of the reference clock and the delay between DCO_output and DCO_clock is measured. A carefully designed reset process reduces the phase acquisition process to two cycles. The ADPLL was implemented using the 32 nm Predictive Technology Model (PTM at 0.9 V supply voltage, and the simulation results show that the proposed ADPLL achieves 10 and 2 reference cycles of frequency and phase acquisitions, respectively, at 700 MHz with less than 67 ps peak-to-peak jitter. The DCO consumes 2.2 mW at 650 MHz with 0.9 V power supply.

  12. Passive Technology to Improve Criticality Control of NTP Reactors, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — This SBIR will develop passive reactor criticality control technology for Nuclear Thermal Propulsion (NTP) identified by Ultra Safe Nuclear Corporation (USNC) in...

  13. Controlling radiation fields in siemans designed light water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Riess, R.; Marchl, T. [Siemens Power Generation Group, Erlangen (Germany)

    1995-03-01

    An essential item for the control of radiation fields is the minimization of the use of satellites in the reactor systems of Light Water Reactors (LWRs). A short description of the qualification of Co-replacement materials will be followed by an illustration of the locations where these materials were implemented in Siemens designed LWRs. Especially experiences in PWRs show the immense influence of reduction of cobalt sources on dose rate buildup. The corrosion and the fatique and wear behavior of the replacement materials has not created concern up to now. A second tool to keep occupational radiation doses at a low level in PWRs is the use of the modified B/Li-chemistry. This is practized in Siemens designed plants by keeping the Li level at a max. value of 2 ppm until it reaches a pH (at 300{degrees}C) of {approximately}7.4. This pH is kept constant until the end of the cycle. The substitution of cobalt base alloys and thus the removal of the Co-59 sources from the system had the largest impact on the radiation levels. Nonetheless, the effectiveness of the coolant chemistry should not be neglected either. Several years of successful operation of PWRs with the replacement materials resulted in an occupational radiation exposure which is below 0.5 man-Sievert/plant and year.

  14. End point control of an actinide precipitation reactor

    International Nuclear Information System (INIS)

    Muske, K.R.

    1997-01-01

    The actinide precipitation reactors in the nuclear materials processing facility at Los Alamos National Laboratory are used to remove actinides and other heavy metals from the effluent streams generated during the purification of plutonium. These effluent streams consist of hydrochloric acid solutions, ranging from one to five molar in concentration, in which actinides and other metals are dissolved. The actinides present are plutonium and americium. Typical actinide loadings range from one to five grams per liter. The most prevalent heavy metals are iron, chromium, and nickel that are due to stainless steel. Removal of these metals from solution is accomplished by hydroxide precipitation during the neutralization of the effluent. An end point control algorithm for the semi-batch actinide precipitation reactors at Los Alamos National Laboratory is described. The algorithm is based on an equilibrium solubility model of the chemical species in solution. This model is used to predict the amount of base hydroxide necessary to reach the end point of the actinide precipitation reaction. The model parameters are updated by on-line pH measurements

  15. Conceptual design of a large Spectral Shift Controlled Reactor

    International Nuclear Information System (INIS)

    Matzie, R.A.; Menzel, G.P.

    1979-08-01

    Within the framework of the Nonproliferation Alternative Systems Assessment Program (NASAP), the US Department of Energy (DOE) has sponsored the development of a conceptual design of a large Spectral Shift Controlled Reactor (SSCR). This report describes the results of the development program and assesses the performance of the conceptual SSCR on the basis of fuel resource utilization and total power costs. The point of departure of the design study was a 1270 MW(e) PWR using Combustion Engineering's System 80/sup TM/ reactor and Stone and Webster's Reference Plant Design. The initial phase of the study consisted of establishing an optimal core design for both the once-through uranium cycle and the denatured U-235/thorium cycle with uranium recycle. The performance of the SSCR was then also assessed for the denatured U-233/thorium cycle with uranium recycle and for the plutonium/thorium cycle with plutonium recycle. After the optimal core design was established, the design of the NSSS and balance of plant was developed

  16. Advanced I&C for Fault-Tolerant Supervisory Control of Small Modular Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Cole, Daniel G. [Univ. of Pittsburgh, PA (United States)

    2018-01-30

    In this research, we have developed a supervisory control approach to enable automated control of SMRs. By design the supervisory control system has an hierarchical, interconnected, adaptive control architecture. A considerable advantage to this architecture is that it allows subsystems to communicate at different/finer granularity, facilitates monitoring of process at the modular and plant levels, and enables supervisory control. We have investigated the deployment of automation, monitoring, and data collection technologies to enable operation of multiple SMRs. Each unit's controller collects and transfers information from local loops and optimize that unit’s parameters. Information is passed from the each SMR unit controller to the supervisory controller, which supervises the actions of SMR units and manage plant processes. The information processed at the supervisory level will provide operators the necessary information needed for reactor, unit, and plant operation. In conjunction with the supervisory effort, we have investigated techniques for fault-tolerant networks, over which information is transmitted between local loops and the supervisory controller to maintain a safe level of operational normalcy in the presence of anomalies. The fault-tolerance of the supervisory control architecture, the network that supports it, and the impact of fault-tolerance on multi-unit SMR plant control has been a second focus of this research. To this end, we have investigated the deployment of advanced automation, monitoring, and data collection and communications technologies to enable operation of multiple SMRs. We have created a fault-tolerant multi-unit SMR supervisory controller that collects and transfers information from local loops, supervise their actions, and adaptively optimize the controller parameters. The goal of this research has been to develop the methodologies and procedures for fault-tolerant supervisory control of small modular reactors. To achieve

  17. Development of a system based in a digital signal processor (DSP) for a simulator of power regulation in a reactor: first stage; Desarrollo de un sistema basado en un DSP para un simulador de regulacion de potencia en un reactor: 1. etapa

    Energy Technology Data Exchange (ETDEWEB)

    Benitez R, J.S.; Perez C, B. [Instituto Nacional de Investigaciones Nucleares, Km. 36.5 Carretera Mexico-Toluca, Municipio de Ocoyoacac, 52045 Estado de Mexico (Mexico)

    2002-07-01

    The first stage of the development of a digital system based on a DSP is presented which forms part of an hybrid simulator for the power regulation in am model of the punctual kinetics of a TRIGA reactor type. The DSP performs the regulation, using a Mandami type algorithm of diffuse control. In the algorithm, the universe of the output variable is discretized for performing in an unique stage the aggregation functions and dis-diffusization. (Author)

  18. Digital control computer upgrade at the Cernavoda NPP simulator

    International Nuclear Information System (INIS)

    Ionescu, T.

    2006-01-01

    The Plant Process Computer equips some Nuclear Power Plants, like CANDU-600, with Centralized Control performed by an assembly of two computers known as Digital Control Computers (DCC) and working in parallel for safely driving of the plan at steady state and during normal maneuvers but also during abnormal transients when the plant is automatically steered to a safe state. The Centralized Control means both hardware and software with obligatory presence in the frame of the Full Scope Simulator and subject to changing its configuration with specific requirements during the plant and simulator life and covered by this subsection

  19. Retrofit of new digital control systems in existing power stations

    International Nuclear Information System (INIS)

    Smith, J.E.; Baird, C.F.

    1986-01-01

    With the notable exception of the Canadian CANDU nuclear power stations, little use has been made of digital control in North American nuclear stations. Recently, however, there has been renewed interest in such systems within the nuclear industry in response to demands for better ergonomics in control room design and the obsolescence of control equipment whose fundamental design has changed little in 20 yr. Early in 1985, Atomic Energy of Canada Limited was asked by New Brunswick Power to advise on the redesign of the control systems for two fossil-fired generating stations, Coleson Cove and Courtenay Bay Unit 4. Coleson Cove is to be converted from oil to coal firing with consequent extensive control system changes and Courtenay Bay Unit 4 required a low-cost solution to the problem of relocating its control room from the existing isolated location to the combined control center used by units 1, 2, and 3. In both cases, the recommended solution involves the retrofit of state-of-the-art digital control systems. Although the units involved are nonnuclear the experience is applicable

  20. Digital Instrumentation and Control working group (DICWG) - MDEP DICWG Programme Plan 2012 2013

    International Nuclear Information System (INIS)

    2012-02-01

    The Multinational Design Evaluation Programme (MDEP) Digital Instrumentation and Controls Working Group (DICWG) was approved by MDEP's Policy Group in March 2008 and meets approximately 3 times a year. All MDEP members and the IAEA are invited to participate in this working group's activities. The DICWG's main objectives are as follows: - to document common positions in the DI and C safety systems design areas; - to harmonise and converge national codes, standards and regulatory requirements and practices in this area while recognising the sovereign rights and responsibilities of national regulators in carrying out their safety reviews of new reactor designs (see the DICWG programme plan for more details of the group's work). The DICWG interacts regularly with the following organisations: - IEC (International Electro-technical Commission) Subcommittee 45A, Instrumentation and Control of Nuclear Facilities; - IEEE (Institute of Electric and Electronics Engineers); - other organisations involved in the design of digital I and C safety systems for nuclear power plants. The DICWG reports its status to the MDEP Steering Technical Committee at the latter's thrice annual meetings. This document presents the 2012 and 2013 programme plan and its products: the Generic Common Position DICWG-02 on Software Tools; the Generic Common Position DICWG-03 on Verification and Validation throughout the Life Cycle of Safety Systems Using Digital Computers; the Generic Common Position DICWG-04 on Communication Independence; the Generic Common Position DICWG-05 on Treatment of Hardware Description Language (HDL) Programmed Devices for Use in Nuclear Safety Systems; the Generic Common Position DICWG-06 on Simplicity in Design; the Generic Common Position DICWG-08 on Impact of Cyber Security Features on Digital I and C Safety Systems

  1. Non-linear model based control of a propylene polymerization reactor

    NARCIS (Netherlands)

    Al-Haj Ali, M.; Betlem, B.; Weickert, G.; Roffel, B.

    2007-01-01

    A modified generic model controller is developed and tested through a simulation study. The application involves model-based control of a propylene polymerization reactor in which the monomer conversion and melt index of the produced polymer are controlled by manipulating the reactor cooling water

  2. The Effect of Degraded Digital Instrumentation and Control systems on Human-system Interfaces and Operator Performance

    Energy Technology Data Exchange (ETDEWEB)

    OHara, J.M.; Gunther, B.; Martinez-Guridi, G. (BNL); Xing, J.; Barnes, V. (NRC)

    2010-11-07

    Integrated digital instrumentation and control (I&C) systems in new and advanced nuclear power plants (NPPs) will support operators in monitoring and controlling the plants. Even though digital systems typically are expected to be reliable, their potential for degradation or failure significantly could affect the operators performance and, consequently, jeopardize plant safety. This U.S. Nuclear Regulatory Commission (NRC) research investigated the effects of degraded I&C systems on human performance and on plant operations. The objective was to develop technical basis and guidance for human factors engineering (HFE) reviews addressing the operator's ability to detect and manage degraded digital I&C conditions. We reviewed pertinent standards and guidelines, empirical studies, and plant operating experience. In addition, we evaluated the potential effects of selected failure modes of the digital feedwater control system of a currently operating pressurized water reactor (PWR) on human-system interfaces (HSIs) and the operators performance. Our findings indicated that I&C degradations are prevalent in plants employing digital systems, and the overall effects on the plant's behavior can be significant, such as causing a reactor trip or equipment to operate unexpectedly. I&C degradations may affect the HSIs used by operators to monitor and control the plant. For example, deterioration of the sensors can complicate the operators interpretation of displays, and sometimes may mislead them by making it appear that a process disturbance has occurred. We used the findings as the technical basis upon which to develop HFE review guidance.

  3. Adaptive slope compensation for high bandwidth digital current mode controller

    DEFF Research Database (Denmark)

    Taeed, Fazel; Nymand, Morten

    2015-01-01

    An adaptive slope compensation method for digital current mode control of dc-dc converters is proposed in this paper. The compensation slope is used for stabilizing the inner current loop in peak current mode control. In this method, the compensation slope is adapted with the variations...... in converter duty cycle. The adaptive slope compensation provides optimum controller operation in term of bandwidth over wide range of operating points. In this paper operation principle of the controller is discussed. The proposed controller is implemented in an FPGA to control a 100 W buck converter....... The experimental results of measured loop-gain at different operating points are presented to validate the theoretical performance of the controller....

  4. Frequency to digital converter for IUAC Linac control system

    International Nuclear Information System (INIS)

    Jain, Mamta; Subramaiam, E.T.; Sahu, B.K.

    2015-01-01

    A frequency to digital converter CAMAC module has been designed and developed for LINAC control systems. This module is used to see the frequency difference of master clock and the resonator frequency digitally without using the oscilloscope. Later on this can be used for automatic tuning and locking of the cavities using piezoelectric actuator based tunner control. This module has eight independent channels to fulfill the need of all the eight cavities of the cryostat. A Schmitt trigger along with level converaccepts almost any form of pulse train, with 30 Vp-p. The time period is measured by counters clocked from a high resolution clock (10 MHz +/- 250 ps). The counter values are cross checked at both the input levels. Frequency is obtained from the computed time period by a special divisor core implemented inside the FPGA. The major task was the implementation of eight individual divisor cores and routing inside one Spartan 3s500E FPGA chip

  5. Digital Control of External Devices through the Parallel Port of a ...

    African Journals Online (AJOL)

    the controlling voltage of 5V) which drives an electrically isolated circuit with a relay. A model of the system was built and the test result was satisfactory. Keywords: device controller, digital switching, digital interfacing, visual basic, computer ...

  6. Sequencing batch-reactor control using Gaussian-process models.

    Science.gov (United States)

    Kocijan, Juš; Hvala, Nadja

    2013-06-01

    This paper presents a Gaussian-process (GP) model for the design of sequencing batch-reactor (SBR) control for wastewater treatment. The GP model is a probabilistic, nonparametric model with uncertainty predictions. In the case of SBR control, it is used for the on-line optimisation of the batch-phases duration. The control algorithm follows the course of the indirect process variables (pH, redox potential and dissolved oxygen concentration) and recognises the characteristic patterns in their time profile. The control algorithm uses GP-based regression to smooth the signals and GP-based classification for the pattern recognition. When tested on the signals from an SBR laboratory pilot plant, the control algorithm provided a satisfactory agreement between the proposed completion times and the actual termination times of the biodegradation processes. In a set of tested batches the final ammonia and nitrate concentrations were below 1 and 0.5 mg L(-1), respectively, while the aeration time was shortened considerably. Copyright © 2013 Elsevier Ltd. All rights reserved.

  7. A powerful ethernet interface module for digital camera control

    Science.gov (United States)

    Amato, Stephen M.; Geary, John C.

    2012-09-01

    We have found a commercially-available ethernet interface module with sufficient on-board resources to largely handle all timing generation tasks required by digital imaging systems found in astronomy. In addition to providing a high-bandwidth ethernet interface to the controller, it can largely replace the need for special-purpose timing circuitry. Examples for use with both CCD and CMOS imagers are provided.

  8. Rapid-L Operator-Free Fast Reactor Concept Without Any Control Rods

    International Nuclear Information System (INIS)

    Kambe, Mitsuru; Tsunoda, Hirokazu; Mishima, Kaichiro; Iwamura, Takamichi

    2003-01-01

    The 200-kW(electric) uranium-nitride-fueled lithium-cooled fast reactor concept 'RAPID-L' to achieve highly automated reactor operation has been demonstrated. RAPID-L is designed for a lunar base power system. It is one of the variants of the RAPID (Refueling by All Pins Integrated Design) fast reactor concept, which enables quick and simplified refueling. The essential feature of the RAPID concept is that the reactor core consists of an integrated fuel assembly instead of conventional fuel subassemblies. In this small-size reactor core, 2700 fuel pins are integrated and encased in a fuel cartridge. Refueling is conducted by replacing a fuel cartridge. The reactor can be operated without refueling for up to 10 yr.Unique challenges in reactivity control systems design have been addressed in the RAPID-L concept. The reactor has no control rod but involves the following innovative reactivity control systems: lithium expansion modules (LEM) for inherent reactivity feedback, lithium injection modules (LIM) for inherent ultimate shutdown, and lithium release modules (LRM) for automated reactor startup. All these systems adopt 6 Li as a liquid poison instead of B 4 C rods. In combination with LEMs, LIMs, and LRMs, RAPID-L can be operated without an operator. This reactor concept is also applicable to the terrestrial fast reactors. In this paper, the RAPID-L reactor concept and its transient characteristics are presented

  9. Some particular aspects of control in nuclear power reactors

    International Nuclear Information System (INIS)

    Vathaire, F. de; Vernier, Ph.; Pascouet, A.

    1964-01-01

    This paper reviews the experience acquired in France on the question, of reactor safety. Since a special paper is being presented on reactors of the graphite gas type, the safety of the other types studied in France is discussed here: - heavy water-gas reactors, - fast neutron reactors, - water research reactors of the swimming-pool and tank types. The safety rules peculiar to the different types are explained, with emphasis on their influence on the reactor designs and on the power limits they impose. The corresponding safety studies are presented, particular stress being placed on the original work developed in these fields. Special mention is made of the experimental systems constructed for these studies: the reactor CABRI, pile loop for depressurization tests, loops outside the pile, mock-ups etc. (authors) [fr

  10. Nuclear reactor control with fuzzy logic approaches - strengths, weakness, opportunities, and threats

    International Nuclear Information System (INIS)

    Ruan, Da

    2004-01-01

    As part of the special track on 'Lessons learned from computational intelligence in nuclear applications' at the forthcoming FLINS 2004 conference on Applied Computational Intelligence (Blankenberge, Belgium, September 1-3, 2004), research experiences on fuzzy logic techniques in applications of nuclear reactor control operation are critically reviewed in this presentation. Assessment of four real fuzzy control applications at the MIT research reactor in the US, the FUGEN heavy water reactor in Japan, the BR1 research reactor in Belgium, and a TRIGA Mark III reactor in Mexico will be examined thought a SWOT analysis (strengths, weakness, opportunities, and threats). Special attention will be paid to the current cooperation between the Belgian Nuclear Research Centre (SCK-CEN) and the Mexican Nuclear Centre (ININ) on the fuzzy logic control for nuclear reactor control project under the partial support of the National Council for Science and Technology of Mexico (CONACYT). (Author)

  11. Estimation of multiply digital process control system extractive distillation stability

    Directory of Open Access Journals (Sweden)

    V. S. Kudryashov

    2016-01-01

    Full Text Available An approach to stability analysis of digital control systems associated non-stationary object on the example of the rectification process. Object modeling with cross-connections and the control scheme of the described system, discrete transfer functions in the shift operators. The equations of connection for each output of the closed-loop system. To solve this problem developed an algorithm for estimating the margin of stability of multivariable digital control systems based on the discrete root criterion, comprising the following main stages: obtaining of the characteristic polynomial of the closed-loop system for each output; computation of eigenvalues of the system matrix in the state space to determine roots of the characteristic equation and the stability of the system; determination of the stability and margin of stability by the deviation of maximum module of the root from the boundary of the high variability. To obtain the characteristic polynomial of a as discrete models of controllers and channels of IP object-use the transfer function of the first order with transport delay. The simulation was performed at different parameters of the control object, which is characterized by a stable and an unstable state of the system. VA-den analysis of the numerical values of the roots and character of their location on the complex plane, which to you-water that the system is stable or unstable. To confirm the obtained results were calculated and presented dynamic characteristics of the closed-loop system under different conditions, which confirm the initial assessment, the root criterion. To determine the factor stability of multivariable digital systems is proposed to use the deviation of the maximum root of the characteristic equation from the stability boundary. The obtained results apply to the class of symmetric multivariable control objects. The approach to assessing the sustainability of multivariable system regulation can be effectively

  12. Design of coordinated controller in nuclear power plant based on digital instrument and control technology

    International Nuclear Information System (INIS)

    Cheng Shouyu; Peng Minjun; Liu Xinkai; Zhao Qiang; Deng Xiangxin

    2014-01-01

    Nuclear power plant (NPP) is a multi-input and multi-output, no-linear and time-varying complex system. The conventional PID controller is usually used in NPP control system which is based on analog instrument. The system parameters are easy to overshoot and the response time is longer in the control mode of the conventional PID. In order to improve this condition, a new coordinated control strategy which is based on expert system and the original controllers in the digital instrument and control technology was presented. In order to verify and validate it, the proposed coordinated control technology was tested by the full-scope real-time simulation system. The results prove that using digital instrument and control technology to achieve coordinated controller is feasible, the coordinated controller can effectively improve the dynamic operating characteristics of the system, and the coordinated controller is superior to the conventional PID controller in control performance. (authors)

  13. Improvement to the control rod drive of a nuclear reactor

    International Nuclear Information System (INIS)

    Desfontaines, Guy.

    1981-01-01

    Improvement to the devices that move the control rods of a nuclear reactor. The slow movements of the rods are generally carried out by screw and nut gear, the nut being blocked as to rotation and the screw as to translation movement. Additionally, a mechanism enables the control rods to be inserted rapidly by release of the screw and nut gear, the nut remaining constantly in gear with the screw. The presence of extra poles and coils under the stator of the actuating motor of the screw add length and weight to the mechanism and hence increase the strains and deformations which affect the latter in the event of an earthquake. The device of the invention makes it possible to overcome this drawback and leads to a more simple mechanism. It is characterized in that the rotor of the motor actuating the screw is also provided with clamps, in its high position, controlled by electromagnetic action as from the coils of the actuating motor stator so that they are in the closed position on the screw when the stator is powered and in the open position when it is no longer so, in order to allow the screw and nut assembly drop, and in that it includes a device to lock the clamps, enabling these to be kept in the open position when the control screw is not in the high holding position [fr

  14. Determination of the control rod worth for research reactors

    International Nuclear Information System (INIS)

    Aldama, D.L.; Gual, M.R.

    2000-01-01

    Nowadays there is a big interest in developing neutronic analysis methods for research reactor and particularly for the determination of the control rods worth under different operation conditions and core configurations. The reactivity associated with the control rods is of interest in the shutdown margin and in calculations of possible abnormal conditions related to reactivity accidents. For theses studies several computer codes have been developed. The present work is aimed at the validation of the calculation methods of the Nuclear Technology Center of Cuba. For this purpose, in order to evaluate the safety of this type of installations, the reactivity worth of the control rods of the cylindrical configuration of the Brazilian critical assembly IPEN/MB-01 is determined. These calculations, however, are a relatively complex task that requires the use of three-dimensional models. Because of this, the validation of the calculation methods used for this purpose is of great importance. In fact, it is one of the requirements called upon by the quality assurance programs for the development, maintenance and utilization of the calculation codes used in safety analysis. For the calculation of control rod worth the lattice code WIMS-D/4 [8] and the diffusion code SNAP-3D [9] were used. This work presents the obtained results and gives a comparison with the experimental values

  15. 47 CFR 73.9001 - Redistribution control of digital television broadcasts.

    Science.gov (United States)

    2010-10-01

    ... 47 Telecommunication 4 2010-10-01 2010-10-01 false Redistribution control of digital television... RADIO SERVICES RADIO BROADCAST SERVICES Digital Broadcast Television Redistribution Control § 73.9001 Redistribution control of digital television broadcasts. Licensees of TV broadcast stations may utilize the...

  16. Experimental computer-controlled instrumentation system for the research reactor DR2

    DEFF Research Database (Denmark)

    Goodstein, L.P.

    1969-01-01

    An instrumentation system has been developed for one of the Danish Atomic Energy Commission's research reactors as part of an experiment on the advantages to be gained by the use of digital computers in a process plant application. Problem areas to be investigated include (a) reliability and safety...

  17. Nuclear reactors

    International Nuclear Information System (INIS)

    Barre, Bertrand

    2015-10-01

    After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

  18. Digital controller for feedwater and recirculation flow control of BWR plant

    International Nuclear Information System (INIS)

    Sato, Takao; Ito, Tetsuo; Omori, Takashi; Iida, Hiroshi; Yanai, Katsuya.

    1980-01-01

    In nuclear power plants, it is required to operate the plants on load-following basis as the proportion of nuclear power plants in whole power supply network has been increasing. For this purpose, the requirements of more reliable, more automated plants and of the flexibility of operation are becoming serious. To respond to such demands, digital controllers are inevitable because analog controllers are limited in their controllability. It is also required to devise more intelligent systems such as those enabling strengthened diagnostic functions or sophisticated predictive control. On such background, duplicated redundant digital control system has been developed, using two control microcomputers and uniting the conventional feed-water control system and recirculation flow control system. The report discribes on the design concept for this digital controller, the hardware and software of the control system and the confirmation of the performance by simulation. The verifying test for the control performance, the simulation test for recirculation pump abnormality, the test for predictive control and the test on the response characteristics of recirculation system were carried out. The digital controller attained the MTBF 10 times as much, and the down time ratio 1/5 as small as those of the analog control systems. (Wakatsuki, Y.)

  19. Research methods of simulate digital compensators and autonomous control systems

    Directory of Open Access Journals (Sweden)

    V. S. Kudryashov

    2016-01-01

    Full Text Available The peculiarity of the present stage of development of the production is the need to control and regulate a large number of process parameters, the mutual influence on each other that when using single-circuit systems significantly reduces the quality of the transition process, resulting in significant costs of raw materials and energy, reduce the quality of the products. Using a stand-alone digital control system eliminates the correlation of technological parameters, to give the system the desired dynamic and static properties, improve the quality of regulation. However, the complexity of the configuration and implementation of procedures (modeling compensators autonomous systems of this type, associated with the need to perform a significant amount of complex analytic transformation significantly limit the scope of their application. In this regard, the approach based on the decompo sition proposed methods of calculation and simulation (realization, consisting in submitting elements autonomous control part digital control system in a series parallel connection. The above theoretical study carried out in a general way for any dimension systems. The results of computational experiments, obtained during the simulation of the four autonomous control systems, comparative analysis and conclusions on the effectiveness of the use of each of the methods. The results obtained can be used in the development of multi-dimensional process control systems.

  20. Stability Analysis and Controller Synthesis for Digital Single-Loop Voltage-Controlled Inverters

    DEFF Research Database (Denmark)

    Wang, Xiongfei; Loh, Poh Chiang; Blaabjerg, Frede

    2016-01-01

    This paper analyzes first the stability of single-loop digital voltage control scheme for the LC-filtered voltage source inverters. It turns out that the phase lag, caused by the time delay of digital control system and by the use of integral controller, can stabilize the voltage loop without...... damping of LC-filter resonance. The stability regions are then identified with alternative voltage controller synthesized. For further widening the stability region, an active damping approach is proposed and co-designed with the voltage controller in the discrete z-domain. Simulations and experimental...

  1. Digitally Controlled Converter with Dynamic Change of Control Law and Power Throughput

    DEFF Research Database (Denmark)

    Nesgaard, Carsten; Andersen, Michael Andreas E.; Nielsen, Nils

    2003-01-01

    the substitution of analog controllers with their digital counterparts are considered. The outline of the paper is divided into two segments – the first being an experimental analysis of the timing behavior by means of code optimization – the second being an examination of the dynamics of incorporating two control......With the continuous development of faster and cheaper microprocessors the field of applications for digital control is constantly expanding. Based on this trend the paper at hand describes the analysis and implementation of multiple control laws within the same controller. Also, implemented within...

  2. Input/Output linearizing control of a nuclear reactor

    International Nuclear Information System (INIS)

    Perez C, V.

    1994-01-01

    The feedback linearization technique is an approach to nonlinear control design. The basic idea is to transform, by means of algebraic methods, the dynamics of a nonlinear control system into a full or partial linear system. As a result of this linearization process, the well known basic linear control techniques can be used to obtain some desired dynamic characteristics. When full linearization is achieved, the method is referred to as input-state linearization, whereas when partial linearization is achieved, the method is referred to as input-output linearization. We will deal with the latter. By means of input-output linearization, the dynamics of a nonlinear system can be decomposed into an external part (input-output), and an internal part (unobservable). Since the external part consists of a linear relationship among the output of the plant and the auxiliary control input mentioned above, it is easy to design such an auxiliary control input so that we get the output to behave in a predetermined way. Since the internal dynamics of the system is known, we can check its dynamics behavior on order of to ensure that the internal states are bounded. The linearization method described here can be applied to systems with one-input/one-output, as well as to systems with multiple-inputs/multiple-outputs. Typical control problems such as stabilization and reference path tracking can be solved using this technique. In this work, the input/output linearization theory is presented, as well as the problem of getting the output variable to track some desired trayectories. Further, the design of an input/output control system applied to the nonlinear model of a research nuclear reactor is included, along with the results obtained by computer simulation. (Author)

  3. 60 GHz 5-bit digital controlled phase shifter in a digital 40 nm CMOS technology without ultra-thick metals

    NARCIS (Netherlands)

    Gao, H.; Ying, K.; Matters-Kammerer, M.K.; Harpe, P.; Wang, B.; Liu, B.; Serdijn, W.A.; Baltus, P.G.M.

    2016-01-01

    A 5-bit digital controlled switch-type passive phase shifter realised in a 40 nm digital CMOS technology without ultra-thick metals for the 60 GHz Industrial, Scientific and Medical (ISM) band is presented. A patterned shielding with electromagnetic bandgap structure and a stacked metals method to

  4. Fuzzy gain scheduling of velocity PI controller with intelligent learning algorithm for reactor control

    International Nuclear Information System (INIS)

    Dong Yun Kim; Poong Hyun Seong; .

    1997-01-01

    In this research, we propose a fuzzy gain scheduler (FGS) with an intelligent learning algorithm for a reactor control. In the proposed algorithm, the gradient descent method is used in order to generate the rule bases of a fuzzy algorithm by learning. These rule bases are obtained by minimizing an objective function, which is called a performance cost function. The objective of the FGS with an intelligent learning algorithm is to generate gains, which minimize the error of system. The proposed algorithm can reduce the time and effort required for obtaining the fuzzy rules through the intelligent learning function. It is applied to reactor control of nuclear power plant (NPP), and the results are compared with those of a conventional PI controller with fixed gains. As a result, it is shown that the proposed algorithm is superior to the conventional PI controller. (author)

  5. Influence of discretization method on the digital control system performance

    Directory of Open Access Journals (Sweden)

    Futás József

    2003-12-01

    Full Text Available The design of control system can be divided into two steps. First the process or plant have to be convert into mathematical model form, so that its behavior can be analyzed. Then an appropriate controller have to be design in order to get the desired response of the controlled system. In the continuous time domain the system is represented by differential equations. Replacing a continuous system into discrete time form is always an approximation of the continuous system. The different discretization methods give different digital controller performance. The methods presented on the paper are Step Invariant or Zero Order Hold (ZOH Method, Matched Pole-Zero Method, Backward difference Method and Bilinear transformation. The above mentioned discretization methods are used in developing PI position controller of a dc motor. The motor model was converted by the ZOH method. The performances of the different methods are compared and the results are presented.

  6. High-Resolution Synthesizable Digitally-Controlled Delay Lines

    Science.gov (United States)

    Giordano, R.; Ameli, F.; Bifulco, P.; Bocci, V.; Cadeddu, S.; Izzo, V.; Lai, A.; Mastroianni, S.; Aloisio, A.

    2015-12-01

    Digitally-controlled delay lines (DCDLs) play a key role in timing distribution for trigger and data acquisition systems (TDAQ) of high energy Physics (HEP), where it is often necessary to add an open-loop fine-grained programmable phase delay to distributed clocks and/or data lines. In this work, we present the performance of DCDLs implemented according to an all-digital novel architecture. The architecture is completely technology-independent, it is described by means of a hardware description language and it can be placed and routed with automatic tools. Our solution is aimed at being used as a synthesizable block in FPGAs, as a proof-of-concept we implemented a prototype in a Xilinx Kintex-7 FPGA. We discuss the measured performance of the implemented delay line in terms of delay range, resolution and linearity. The logic utilization of the delay lines is also presented in the view of a scalable implementation.

  7. Speed Digital Control of Brushless DC Motor Using dsPIC Controller

    Directory of Open Access Journals (Sweden)

    Gheorghe Băluţă

    2014-09-01

    Full Text Available This paper presents the digital control of the Brushless DC motor (BLDCM speed. The dsPICDEM MC1 development system (with the dsPIC30F6010A microcontroller and the dsPICDEM MC1L power module, manufactured by Microchip Company, were used. The control program was developed in C programming language. The graphical user interface was realized in LabVIEW 8.6 graphical programming language. For speed control, a digital controller PI type was implemented. Due to digital controller well chosen and well tuned, the system response at speed step variation is very good. Therewith, the experimental results obtained also show a good compensation of disturbance which does not happen in open-loop control.

  8. Control parameter optimization for AP1000 reactor using Particle Swarm Optimization

    International Nuclear Information System (INIS)

    Wang, Pengfei; Wan, Jiashuang; Luo, Run; Zhao, Fuyu; Wei, Xinyu

    2016-01-01

    Highlights: • The PSO algorithm is applied for control parameter optimization of AP1000 reactor. • Key parameters of the MSHIM control system are optimized. • Optimization results are evaluated though simulations and quantitative analysis. - Abstract: The advanced mechanical shim (MSHIM) core control strategy is implemented in the AP1000 reactor for core reactivity and axial power distribution control simultaneously. The MSHIM core control system can provide superior reactor control capabilities via automatic rod control only. This enables the AP1000 to perform power change operations automatically without the soluble boron concentration adjustments. In this paper, the Particle Swarm Optimization (PSO) algorithm has been applied for the parameter optimization of the MSHIM control system to acquire better reactor control performance for AP1000. System requirements such as power control performance, control bank movement and AO control constraints are reflected in the objective function. Dynamic simulations are performed based on an AP1000 reactor simulation platform in each iteration of the optimization process to calculate the fitness values of particles in the swarm. The simulation platform is developed in Matlab/Simulink environment with implementation of a nodal core model and the MSHIM control strategy. Based on the simulation platform, the typical 10% step load decrease transient from 100% to 90% full power is simulated and the objective function used for control parameter tuning is directly incorporated in the simulation results. With successful implementation of the PSO algorithm in the control parameter optimization of AP1000 reactor, four key parameters of the MSHIM control system are optimized. It has been demonstrated by the calculation results that the optimized MSHIM control system parameters can improve the reactor power control capability and reduce the control rod movement without compromising AO control. Therefore, the PSO based optimization

  9. Trajectory-tracking control of underwater inspection robot for nuclear reactor internals using Time Delay Control

    Energy Technology Data Exchange (ETDEWEB)

    Park, Joon-Young [Green Growth Laboratory, Korea Electric Power Research Institute, 65 Munjiro, Yuseong-Gu, Daejeon (Korea, Republic of)], E-mail: asura@kepco.co.kr; Cho, Byung-Hak; Lee, Jae-Kyung [Green Growth Laboratory, Korea Electric Power Research Institute, 65 Munjiro, Yuseong-Gu, Daejeon (Korea, Republic of)

    2009-11-15

    This paper addresses the trajectory control problem of an underwater inspection robot for nuclear reactor internals. From the viewpoint of control engineering, the trajectory control of the underwater robot is a difficult task due to its nonlinear dynamics, which includes various hydraulic forces such as buoyancy and hydrodynamic damping, the difference between the centres of gravity and buoyancy, and disturbances from a tether cable. To solve such problems, we applied Time Delay Control to the underwater robot. This control law has a very simple structure not requiring nonlinear plant dynamics, and was proven to be highly robust against nonlinearities, uncertainties and disturbances. We confirmed its effectiveness through experiments.

  10. Trajectory-tracking control of underwater inspection robot for nuclear reactor internals using Time Delay Control

    International Nuclear Information System (INIS)

    Park, Joon-Young; Cho, Byung-Hak; Lee, Jae-Kyung

    2009-01-01

    This paper addresses the trajectory control problem of an underwater inspection robot for nuclear reactor internals. From the viewpoint of control engineering, the trajectory control of the underwater robot is a difficult task due to its nonlinear dynamics, which includes various hydraulic forces such as buoyancy and hydrodynamic damping, the difference between the centres of gravity and buoyancy, and disturbances from a tether cable. To solve such problems, we applied Time Delay Control to the underwater robot. This control law has a very simple structure not requiring nonlinear plant dynamics, and was proven to be highly robust against nonlinearities, uncertainties and disturbances. We confirmed its effectiveness through experiments.

  11. Extraction of gadolinium from high flux isotope reactor control plates

    International Nuclear Information System (INIS)

    Kohring, M.W.

    1987-04-01

    Gadolinium-153 is an important radioisotope used in the diagnosis of various bone disorders. Recent medical and technical developments in the detection and cure of osteoporosis, a bone disease affecting an estimated 50 million people, have greatly increased the demand for this isotope. The Oak Ridge National Laboratory (ORNL) has produced 153 Gd since 1980 primarily through the irradiation of a natural europium-oxide powder followed by the chemical separation of the gadolinium fraction from the europium material. Due to the higher demand for 153 Gd, an alternative production method to supplement this process has been investigated. This process involves the extraction of gadolinium from the europium-bearing region of highly radioactive, spent control plates used at the High Flux Isotope Reactor (HFIR) with a subsequent re-irradiation of the extracted material for the production of the 153 Gd. Based on the results of experimental and calculational analyses, up to 25 grams of valuable gadolinium (≥60% enriched in 152 Gd) resides in the europium-bearing region of the HFIR control components of which 70% is recoverable. At a specific activity yield of 40 curies of 153 Gd for each gram of gadolinium re-irradiated, 700 one-curie sources can be produced from each control plate assayed

  12. A Digitally Controllable Polymer-Based Microfluidic Mixing Module Array

    Directory of Open Access Journals (Sweden)

    Raymond H. W. Lam

    2012-03-01

    Full Text Available This paper presents an integrated digitally controllable microfluidic system for continuous solution supply with a real-time concentration control. This system contains multiple independently operating mixing modules, each integrated with two vortex micropumps, two Tesla valves and a micromixer. The interior surface of the system is made of biocompatible materials using a polymer micro-fabrication process and thus its operation can be applied to chemicals and bio-reagents. In each module, pumping of fluid is achieved by the vortex micropump working with the rotation of a micro-impeller. The downstream fluid mixing is based on mechanical vibrations driven by a lead zirconate titanate ceramic diaphragm actuator located below the mixing chamber. We have conducted experiments to prove that the addition of the micro-pillar structures to the mixing chamber further improves the mixing performance. We also developed a computer-controlled automated driver system to control the real-time fluid mixing and concentration regulation with the mixing module array. This research demonstrates the integration of digitally controllable polymer-based microfluidic modules as a fully functional system, which has great potential in the automation of many bio-fluid handling processes in bio-related applications.

  13. Safety and position control system for a nuclear reactor control rod

    International Nuclear Information System (INIS)

    Jacquelin, Roland; Michot, Gilbert.

    1982-01-01

    This device comprises a vertically mobile tube, terminating at its bottom end with an electromagnet maintaining the control rod, and of which the upper end is maintained by a second electromagnet, so that when the current to the two electromagnets is cut simultaneously the tube drops under the effect of gravity, thereby helping with its weight to push the control rod into its sleeve, even if the latter has accidental distortions. Application is for nuclear reactors [fr

  14. Research reactor instrumentation and control technology. Report of a technical committee meeting

    International Nuclear Information System (INIS)

    1997-10-01

    The majority of research reactors operating today were put into operation 20 years ago, and some of them underwent modifications, upgrading and refurbishing since their construction to meet the requirements for higher neutron fluxes. However, a few of these ageing research reactors are still operating with their original instrumentation and control systems (I and C) which are important for reactor safety to guard against abnormal occurrences and reactor control involving startup, shutdown and power regulation. Worn and obsolete I and C systems cause operational problems as well as difficulties in obtaining replacement parts. In addition, satisfying the stringent safety conditions laid out by the nuclear regulatory bodies requires the modernization of research reactors I and C systems and integration of additional instrumentation units to the reactor. In order to clarify these issues and to provide some guidance to reactor operators on state-of-art technology and future trends for the I and C systems for research reactors, a Technical Committee Meeting on Technology and Trends for Research Reactor Instrumentation and Controls was held in Ljubljana, Slovenia, from 4 to 8 December 1995. This publication summarizes the discussions and recommendations resulting from that meeting. This is expected to benefit the research reactor operators planning I and C improvements. Refs, figs, tabs

  15. A reconfigurable strategy for distributed digital process control

    International Nuclear Information System (INIS)

    Garcia, H.E.; Ray, A.; Edwards, R.M.

    1990-01-01

    A reconfigurable control scheme is proposed which, unlike a preprogrammed one, uses stochastic automata to learn the current operating status of the environment (i.e., the plant, controller, and communication network) by dynamically monitoring the system performance and then switching to the appropriate controller on the basis of these observations. The potential applicability of this reconfigurable control scheme to electric power plants is being investigated. The plant under consideration is the Experimental Breeder Reactor (EBR-II) at the Argonne National Laboratory site in Idaho. The distributed control system is emulated on a ring network where the individual subsystems are hosted as follows: (1) the reconfigurable control modules are located in one of the network modules called Multifunction Controller; (2) the learning modules are resident in a VAX 11/785 mainframe computer; and (3) a detailed model of the plant under control is executed in the same mainframe. This configuration is a true representation of the network-based control system in the sense that it operates in real time and is capable of interacting with the actual plant

  16. Development of a digital card to simulate period transients in research reactors

    International Nuclear Information System (INIS)

    Masotti, Paulo Henrique Ferraz

    1999-01-01

    This work presents the development of a card to be used in a 'slot' of a micro-computer for evaluation of a nuclear channel used to monitor the start up of nuclear reactors. The results of the bench tests showed good linearity and 2% error deviation in the entire range of operation. Fields tests, performed with the start up channel of IEA-R1 research reactor showed that the card is an excellent device to verify the performance of the channel during steady state, and transient conditions. (author)

  17. Digital Signal Processing and Control for the Study of Gene Networks

    Science.gov (United States)

    Shin, Yong-Jun

    2016-04-01

    Thanks to the digital revolution, digital signal processing and control has been widely used in many areas of science and engineering today. It provides practical and powerful tools to model, simulate, analyze, design, measure, and control complex and dynamic systems such as robots and aircrafts. Gene networks are also complex dynamic systems which can be studied via digital signal processing and control. Unlike conventional computational methods, this approach is capable of not only modeling but also controlling gene networks since the experimental environment is mostly digital today. The overall aim of this article is to introduce digital signal processing and control as a useful tool for the study of gene networks.

  18. Digital Signal Processing and Control for the Study of Gene Networks.

    Science.gov (United States)

    Shin, Yong-Jun

    2016-04-22

    Thanks to the digital revolution, digital signal processing and control has been widely used in many areas of science and engineering today. It provides practical and powerful tools to model, simulate, analyze, design, measure, and control complex and dynamic systems such as robots and aircrafts. Gene networks are also complex dynamic systems which can be studied via digital signal processing and control. Unlike conventional computational methods, this approach is capable of not only modeling but also controlling gene networks since the experimental environment is mostly digital today. The overall aim of this article is to introduce digital signal processing and control as a useful tool for the study of gene networks.

  19. Mechanism design for the control rods conduction of TRIGA Mark III reactor in the NINR

    International Nuclear Information System (INIS)

    Franco C, A.

    1997-01-01

    This work presents in the first chapter a general studio about the reactor and the importance of control rods in the reactor , the mechaniucal design attending to requisitions that are imposed for conditions of operation of the reactor are present in the second chapter, the narrow relation that exists with the new control console and the mechanism is developed in the thired chapter, this relation from a point of view of an assembly of components is presents in fourth chapter, finally reaches and perspectives of mechanism forming part of project of the automation of reactor TRIGA MARK III, are present in the fifth chapter. (Author)

  20. Investigating The Integral Control Rod Worth Of The Miniature Neutron Source Reactor MNSR

    International Nuclear Information System (INIS)

    Nguyen Hoang Hai; Do Quang Binh

    2011-01-01

    Determining control rod characteristics is an essential problem of nuclear reactor analysis. In this research, the integral control rod worth of the miniature neutron source reactor MNSR is investigated. Some other parameters of the nuclear reactor, such as core excess reactivity, shut down margin, are also calculated. Group constants for all reactor components are generated by the WIMSD code and then are used in the CITATION code to solve the neutron diffusion equations. The maximum relative error of the calculated results compared with the measurement data is about 3.5%. (author)